IR 05000346/1998005
| ML20249A607 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/11/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20249A604 | List: |
| References | |
| 50-346-98-05, 50-346-98-5, NUDOCS 9806170136 | |
| Download: ML20249A607 (21) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 111 Docket No: 50-346 License No: NPF-3 Report No: 50-346/98005(DRP)
l Licensee: Toledo Edison Company J l
Facility: Davis-Besse Nuclear Power Station ]
Location: 5501 N. State Route 2 ;
Oak Harbor, OH 43449 ;
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Dates: April 1 - May 12,1998 .
l Inspectors: S. Campbell, Senior Resident inspector l K. Zellers, Resident inspector
Approved by: Thomas J. Kozak, Chief Reactor Projects Branch 4
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9906170136 990611
[' . PDR ADOCK 05000346 o G PDR
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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Station NRC Inspection Report 50-346/98005(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspectio Operations
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Operators reacted very well to an unexpected resin burst in the letdown system that plugged filters in the system and caused difficulty in controlling pressurizer level. The operators prevented resin from entering the reactor coolant system (RCS) and
, appropriately tripped the reactor before the increasing water level could challenge the pressurizer code safety valves (Section 01.1).
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The licensee appropriately considered shutdown risk before scheduling activities affecting the operation of the decay heat removal system and during reactor vessel draining activities. Generally, the operators used good communication techniques and had appropriate control of the drain down evolution. Procedures regarding reduced inventory and mid-loop operations met the intent of Generic Letter 88-17, " Loss of Shutdown
= Cooling"(Sections 01.2 and O1.3).
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Poor communication among control room personnel regarding a redundant valve controller failure occurred while lowering reactor vessel water level (Section 01.3).
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During the transfer of radioactive water from the reactor coolant drain tank to the clean waste receiver tank, a degraded filter inlet isolation valve caused the water to be diverted to the miscellaneous waste tank. Operators did not recognize the water diversion until approximately 8000 gallons had been transferred (Section 01.3).
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A procedural weakness was identified in that a formal RCS inventory balance was not performed. This weakness could have been identified earlier if two information notices had been dispositioned more aggressively (Section 01.3).
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The inspectors concluded the licensee effectively prepared for outage activities by scripting risk significant tasks (Section 06.1).
- The establishment of a work support cerater outside of the control room reduced the distractions and the administrative tasks that operators in the control room had previously been required to perform. The inspectors confirmed that the operators remained aware of plant activities (Section 06.2).
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During main steam safety valve testing, valves found out-of-tolerance were reset. Even though the main steam headers remained operable, the licensee used the guidance of i
NUREG-1022, " Event Reporting Guidelines for 10 CFR 50.72 and 50.73," to report an unusual number of valves outside setpoint tolerances (Section M1.2).
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Several near-misses involving crane and rigging operations occurred in containment that jeopardized plant equipment and personnel safety. The licensee was addressing the collective significance of this issue (Section M1.3).
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The inspectors concluded that refueling activities were performed in accordance with administrative procedures that provided multiple barriers to ensure that fuel movements were performed as planned (Section M1.5).
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Two light fixtures were found broken which resulted in broken glass in the refueling canal and the spent fuel pool. This may have been prevented had the actions taken to address a 1996 audit finding been more comprehensive (Section M2.1).
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The procedure used to fill the steam generator using a temporary manifold was not adequate to ensure all manifold valves used for this activity were controlled (Section M3.1).
Enaineerina
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' The inspectors concluded that engineering personnel conducted a comprehensive inspection and evaluation of the effects of dust from sandblasting the shield building annulus en equipment in the affected rooms (Section E2.2).
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The inspectors observed the meetings of the Station Review Board and Management Review Committee; and evaluated the activities of Quality Assessment personne Overall, these organizations and individuals were effective in their respective roles (Section E7.1).
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Report Details
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Summary of Plant Status l
At the beginning of the inspection, the plant was at 79 percent power and coasting down in power due to depleted fuel. On April 10, during the normal reactor shutdown for the refueling outage, the reactor was manually tripped at 31 percent power after a letdown system demineralized failed. Following the trip, the licensee commenced the eleventh refueling outage (RF-11).
I. Operations 01 Conduct of Operations 0 Manual Reactor Trio Durino Unit Shutdown Insoection Scope (93702)
The inspectors reviewed the licensee's actions prior and subsequent to the manual trip of the reactor due to a failed purification system demineralize Observations and Findinos At 9:45 p.m., on April 10, Makeup Filter #2 became plugged with resin discharged from Purification Demineralized 1-3 that the operators had placed in service to remove lithium from the reactor coolant system (RCS). In response to the plugged filter, the operators removed Makeup Filter #2 from service and placed Makeup Filter #1 in service per Procedure DB-OP-02002," Letdown and Makeup Alarm Procedure." When resin also plugged Makeup Filter #1, letdown pressure increased above the 150 psig setpoint of Purification System Relief Valve PSV-1890. The high pressure caused the valve to open Gnd relieve water to the reactor coolant drain tank, in response to the opened relief valve, the operators entered Procedure DB-OP-02522,"Small RCS Leaks." The operators isolated letdown and reduced sealinjection flow to the reactor coolant pumps from 32 g.p.m. to 12 7p.m. One running makeup pump provided the sealinjection flow and provided a 15 g.p.m. bypass flow around a closed makeup isolation valve. The total makeup flow to the RCS caused pressurizer level to increase.
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At 10:00 p.m., as pressurizer level continued to increase, the shift supervisor decided to begin a rapid shutdown of the reactor per Procedure DB-OP- 02505, " Rapid Shutdown."
At 10:02 p.m., operators reestablished the letdown flow and tried to reduce pressurizer level by diverting flow to the clean waste receiver tanks. However, these efforts were unsuccessful because resin also had plugged filters to the clean waste receiver tank. At 10:32 p.m., the plant was manually tripped from 31 percent power when pressurizer level approached the Technical Specification (TS) 3.4.4 limit of 305 inches. The licensee initiated potential condition adverse to quality report (PCAQR) 98-0529 documenting the reactor trip and the problems with plugged purification system filter The inspectors observed post-trip operator actions and reviewed post-trip data and log Both the integrated control system (ICS) and the operators controlled RCS pressure, RCS l
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temperature and steam generator levels within nominal parameters. The inspectors verified that plant components and systems generally responded as designed. However, Letdown System Control Valve MU-4 failed to shut completely, which was compensated '
for by an upstream valve. Further, one source range instrument failed to energize immediately. Also, the ICS control for one of the main feed water control valves was sluggish which resulted in a lower than expected steam generator pressure. This was attributed to the fact that the ICS was calibrated assuming the plant had been operating at 100 percent power for an extended period rather than 31 percent power, the power level when the plant trippe The inspectors reviewed the procedures the operators had used during the event, the data acquisition and display system (DADS) plots for pressurizer level, and the control room logs covering the period of the event. The inspectors also conducted interviews with the members of the operating crew to verify operator actions taken. The inspectors -
determined that the operators' actions were in accordance with the appropriate
procedures and that no TS requirements were violated during the even Conclusions Operators reacted very well to an unexpected resin burst in the letdown system that plugged filters in the system and caused difficulty in controlling pressurizer level. The
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operators appropriately tripped the reactor as TS limits for pressurizer level were reached. Operators responded to the event in accordance with appropriate operating and abnormal procedures. Plant equipment generally responded as designed during and after the plant tri .2 Schedulina of Work Activities Affectina Decav Heat Removal While in Reduced Inventory Inspection Scope (71707)
The inspectors followed up on a decrease in decay heat removal (DHR) pump suction flow during testing and log on entries regarding midscale failures of control room indicators following replacement of a Nonnuclear instrumentation (NNI) modul Observations and Findinas-Inadvertent Loss Of Nonnuclear instrumentation (NNI) Channel X
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the new module was inserted, the module pins did not appropriately align and a short circuit occurred which resulted in a partialloss of the NNI-X power supply. The power loss caused several midscale failures of control room indicators and the capability to 1 operate Decay Heat Cooler #2 inlet and outlet valves remotely was also lost. At the time j of the failure, the plant was in a reduced inventory condition with reactor vessel level at
'80 inches with DHR Train #2 ir' service, in response to the failure, the operators entered Abnormal Procedure DB-OP-02532, " Loss of NN1/ ICS Power," and verified that incore temperature and decay heat flow had not changed. Control room indicators and NNI power were restored when the module was re-inserted properl !
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Stroking of DHR Suction Valve At 1:19 a.m. on April 17, while the plant wae in a reduced inventory condition, the licensee repaired a packing leak on DHR Pun?o #2 Suction Valve DH-11 by tightening the packing nut. Before stroking the valve for testia q, the operators opened bypass valves to maintain DHR flow to the core. While the operaters stroked the valve closed, suction pressure momentarily dropped below the minimum suction pressure curve for DHR flow in Procedure DB-PF-06703, " Miscellaneous Operat'on Curves." The operators quickly decreased decay heat removal flow to increase suction pressure. The pump never cavitated and DHR flow was always maintained during the drop in pressure. The licensee improved the procedure to clarify that DHR flow be increased to prevent a drop in suction pressure during future valve stroking activitie The inspectors questioned whether replacing the module and stroking Valve DH-11 while in a reduced inventory condition was prudent given the vulnerability to losing DHR capability. The licensee stated that conducting these activities in a reduced inventory condition had minimal risk. The inspectors compared Generic Letter 88-17, " Loss of Shutdown Cooling," guidance with the requirements in Procedure DB-OP-06904,
" Shutdown Operation," and determined that the licensee met the generic letter requirements for controlling activities while the plant was in a reduced inventory conditio The licensee stated that an evaluation would be conducted to determine if rescheduling these activities to a time when the RCS was not in a reduced inventory condition in future outages would be pruden Conclusion The licensee appropriately considered shutdown risk before scheduling two activities (NNI module replacement and stroking DHR Pump #2 suction valve) that could have impacted DHR capability. The licensee met the intent of Generic Letter 88-17 " Loss of Shutdown Cooling" for controlling activities while in reduced inventor .3 RCS Drainina Inspection Scope (71707)
The inspc.ctors observed portions of draining the RCS to mid-loop level (18 inches) for steam generator nozzle dam installatio Observations and Findinas Overall Observation and Assessment During RCS Draining Operations personnel displayed good command and control during the initial draining of the RCS level to mid-loop. The pre-evolution briefs communicated relevant operating experiences from other facilities and necessary responses to unexpected condition During drain down, shift manning was augmented so that critical parameters, such as DHR pump suction pressure and reactor coolant system water levelindication, were observed by dedicated licensed reactor operators. The control room staff was sufficiently staffed with reactor operators monitoring critical parameters and senior reactor operators fulfilling various operation's management oversight role _ _ _ _ _ _ _ _ .
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In a separate event that occurred during RCS draining, operator response to demineralized water falling on top of the operating DHR pump (See Section M3.1) was good in that RCS draining was halted to evaluate the situation. Further, RCS draining I was immediately stopped when level indication was determined to be inconsistent, until a kink in the redundant tygon tube level indication system was found and correcte The inspectors observed that operations personnel followed procedures and implemented good shutdown risk control. As part of the risk control, operators tracked the time for water to boil in the core while at mid-loop level and posted work control signs at the access points of protected equipment trains. The inspectors reviewed the licensee's response and commitment to the recommendations of Generic Letter 88-17," Loss of Decay Heat Removal." The inspectors confirmed that the commitments had been implemented in operating and shutdown risk control procedure Poor Communication of a Failed Redundant Train Valve After draining the reactor from 30 to 18 inches, the inspectors observed that two operators did not communicate that the controller for the non-operating (but protected)
DHR train Heat Exchanger Outlet Valve DH-148, had failed during the draining. Although the DHR lineup procedure instructions to address a controller failure were followed, the shift supervisor and the oversight manager said that they would have stopped the draining evolution had they been aware of the failure. Following the evolution, the outlet valve was verified to be operable by opening it with its safety features actuation system control switch. Additionally, a contingency plan was developed for operation of the DHR train with the valve in the full open positio Inadvertent Diversion of Radioactive Water and Weakness with Reactor Coolant inventory Balance On April 14, the licensee drained RCS water from the reactor vessel to the reactor coolant drain tank (RCDT). The draining activity required that the water in the RCDT be transferred to the clean waste receiver tank (CWRT). Primary Demineralized Filter 1-1 was in the flow path between the RCDT and the CWRT. While transferring the water to the CWRT, a closed inlet filter valve for Primary Demineralized Filter 1-1 which had been removed from service, leaked water to the filter room and the floor drains diveried the water to the CWRT sump. The water in the sump was automatically pumped to the miscellaneous waste tank (MWT).
After several hours, an equipment operator touring the auxiliary building noticed the water draining to the CWRT sump and notified control room personnel. In response to the notification, the control room operators halted reactor vessel draining and the transfer of water from the RCDT to the CWRT. These actions stopped the diversion of water to the sump. More than 8000 gallons of water had been unintentionally diverted to the MW No personnel were contaminated because of the event, but radiation levels in some areas in the auxiliary building did increase. The licensee cleaned the areas, replaced the filter and restored the equipment before restarting the RCS draining evolution. This event was documented on PCAQR 98-057 Operations management recognized that they did not correlate the water volume removed from the RCS with the volume collected in the CWRT. Instead, the operators
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compared the rate of RCS volume change with the change in RCDT level. Consequently, the licensee developed a spreadsheet to track RCS volume changes with CWRT volume changes. The inspectors confirmed this method was accurate to within 1000 gallon However, had CWRT volume changes been tracked initially, the amount of water unintentionally diverted to the MWT would have been detected much earlier with less radiological consequenc The licensee effectively tracked reactor vessel inventory during the drain down evolutio However, the licensee had prior opportunities to develop detailed procedural guidance for an RCS water inventory balance. One issue discussed in Information Notice (lN) 96-65,
" Undetected Accumulation of Gas in Reactor Coolant System and inaccurate Reactor Water Level Indication During Shutdown," pedained to a licensee that did not have detailed procedural guidance requiring an inventory balance for draining evolutions. The licensee dispositioned this issue noting that operator training reinforced the necessity of i'
investigating unexplained changes in RCS inventory. Consequently, the licensee did not
< recommend a change to the procedure to provide more guidance. Additionally, IN 96-37,
" inaccurate Reactor Water Level Indication and inadvertent Drain Down During Shutdown," stated that although tank levels were noted three times a day, inventory balances were not determined from the data. The answer to PCAQR 96-1016, written in response to IN 96-37, stated that station procedures contained adequate procedural requirements to ensure vent paths while in the drained condition. The licensee failed to evaluate whether inventory balances were procedurally require Conclusions Generally, the inspectors noted that operators had good communication and control of j the drain down evolution. It was performed in accordance with procedures that reflected I a careful consideration of shutdown risk and Generic Letter 88-17, " Loss of Shutdown Cooling," recommendations. Poor communication among control room personnel conceming a redundant valve controller failure prevented management from evaluating q the condition. During the transfer of radioactive water from the RCDT to the CWRT, a l degraded filter inlet isolation valve caused the water to be diverted to the MW Operators did not recognize the water diversion until approximately 8000 gallons had been transferre !
06 Operations Organization and Administration ,
0 Plannina and Preparation for the Refuelina Outaae
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The licensee wrote scripts for doing specific risk significant outage activities such as the !
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reactor shutdown and cooldown, primary and secondary system draining, and integrated safety features actuation system testing. The scripting involved the selection of key i outage project managers to be responsible for each task. Tasks were loaded into a !
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computer and a time line of the activities was generated to identify potential conflicts in evolutions. Managers developed task books from existing plant procedures and assigned
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individuals the responsibility for doing a task. The individuals used the task books to walk I down the activity and to setup materials as needed before the activity began. The l inspectors concluded the licensee effectively prepared for outage activities by scripting risk sigr,ificant task l l
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06.2 Use of Operations Work Support Center Durina the Refuelino Outaae (71707)
During previous outages, all outage work packages were authorized by operators in the control room. During this outage, the inspectors noted that this task was shifted to the operations work support center, which was recently established outside the control roo This resulted in a significant improvement to the control room environment. Additionally, moving the work-related activities out of the control room provided a better environment for maintenance related discussions, briefs, and interfaces to take place. Operators remained aware of activities occurring in the plan II. Maintenance M Conduct of Maintenance M1.1 Maintenance and Surveillance Activities (61726 and 62707)
The following maintenance and surveillance testing activities were observed / reviewed during the inspection period:
- DB-MM-03001 Main Steam Safety Valve Setpoint Test
~ DB-PF-03771 Core Flood Tank 2 Check Valve Test
- DB-PF-03298 HPl Check Valve Forward Flow Test Train 2 MWO-3-98-4519-01 Weld Repair of Main Steam Isolation Valve MS-100 MWO-3-96-0699-01 Preventive Maintenance on YV2 Inverter and YRF2 Rectifier MWO-3-98-3898-01 Replace Solenoid Valve SVMU66A MWO 1-96-0647-00 FEHP4B - Clean / Weld Plugs MWO-3-98-4654-01 MS-734 Check Valve Inspection Problems with maintenance work order (MWO) implementation were brought to the attention of work control management for resolution. Cleanliness, radiological, procedural, tagout and post-maintenance testing requirements were clearty documented in the work packages. Work control management displayed effective control over work
. status and mode restraint checklist M1.2 Setooint Testina of Main Steam Safetv Valves (61726)
p During setpoint testing of main steam safety valves, five safety valves on main Steam Header #1 and three safety valves on main Steam Header #2 lifted at values less than the setpoint tolerance allowed by TS. Two valves lifted at values greater than the setpoint tolerance on main Steam Header #2. The licensee initiated PCAQRs 98-0504,
-0510, and -0521 to document that the valves were discovered outside of TS toleranc The safety valves were adjusted, retested, and determined to be within specification.
l The inspectors reviewed TS 3.7.1.1 and concluded that the licensee met the TS action requirements for valves outside setpoint tolerance. Nevertheless, the licensee used the guidance in NUREG-1022, Revision 1, " Event Reporting Guidelines 10 CFR 50.72 and 50.73," and issued LER 1998-001 to report to the NRC the multiple valves found out-of-specificatio l
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M1.3 Crane Uft Control Inspection Scope (71707)
The inspectors conducted a review of the events that occurred during the use of the containment cran Observations and Findinas Crane Control Several events occurred involving crane operations that jeopardized plant equipment and personnel safety. On April 13, a wire support cable for the polar crane control pendant broke and caused a 100-pound pendant with cabling of several hundred pounds to fall about 140 feet nearly missing personnel below. On April 18, the power cable for the reactor service crane broke inside the cable reel when the cable became entangled on scaffolding pipe constructed over the reactor service crane rails. On April 19, a jib arm on the polar crane trolley hit a winch cable supporting a ball and hook rigging device. The device fell approximately 200 feet into the shallow end of the refueling canal missing personnel by 3 feet. On April 21, a fight fixture was knocked off its mount by the reactor service crane control pendant as the crane operator was walking the pendant in the refueling canal area. The light fixture bulb fell into the refueling canal and shattere Lift of Reactor Coolant Pump Ratchet Plate Over the Refueling Canal On April 30, a crane operator used the polar crane and lifted a 2000 pound rachet plate for the reactor coolant motor over the refueling canal during fuel handling operations. A nuclear engineer involved in fuel handling activities below noted the inappropriate lift path and checked that the rachet plate was not lifted over any fuel assemblie Procedure DB-MM-06002, " Polar Crane Operation," prohibited lifting of items not related to reactor maintenance, inspection and refueling over the refueling canal during fuel handling operations. Consequently, refueling was stopped and the involved crane operator and other crane operators were counseled regarding the procedural requirements. The licensee reviewed training records and interviewed crane operators and confirmed that all of the crane operators had received adequate training and were knowledgeable of the load path restrictions. The inspectors concluded that these immediate corrective actions were acceptable. The failure to follow Procedure OB-MM-06002 by lifting the rachet plate for the reactor coolant pump motor over the refueling canal during fuel handling operations is a violation of TS 6. Although a violation was issued for an unauthorized heavy lift over the reactor vessel during the last refueling outage, the inspectors concluded that the root causes for that violation and the violation identified during this inspection were different and thus, it was not reasonable to expect the corrective actions for the previously identified violation to prevent the more recent violation. This non-repetitive, licensee-identified and corrected violation is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-346/98005-01(DRP)).
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Lift Control The licensee identified near misses involving crane and rigging operations due to inadequate rigging and storage of rigging equipment. On April 16, the rigging from eddy current equipment came loose, dropping the equipment 15 to 20 feet, almost striking workers who were erecting scaffolding below. Additionally, a loose chain above Reactor Coolant Pump 2-2 fell through a floor opening and nearly missed workers. Finally, a portable transformer fell 5 feet out of a cart while being lifted, nearly hitting personne No personnelinjuries or damage to equipment occurred because of these events. Each event was documented with a PCAQR. Further, the managemere review committee generated PCAQR 98-0848 to evaluate the collective significance of events pertaining to crane operations and rigging within containment. Station management had also expressed concem over the number of events and had required crane operator stand downs to increase awareness of the requirements and precautions to take while conducting containment crane operation Conclusions Several significant near-misses involving crane and rigging operations occurred in containment that jeopardized plant equipment and personnel safety. The licensee was addressing the collective significance of this issu M1.5 Fuel Handlina Operations (6270_ The inspectors reviewed procedures DB-NE-06101, " Fuel / Component Shuffle," and i
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DB-OP-00030 " Fuel Handling Operations," and determined that the procedures provided multiple independent verifications and tracking mechanisms to ensure that the correct l
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fuel assembly was placed into its designated position in the either the core or the spent fuel pool. The inspectors observed fuel handling activities and determined that personnel followed the procedures. Refueling was stopped when procedure problems were discovered during the activity and were resolved before proceeding with refueling. The inspectors concluded that refueling activities were performed in accordance with administrative procedures that provided multiple barriers to ensure that fuel movements would be performed as planne M2 - Maintenance and Material Condition of Facilities and Equipment M Broken Glass in the Refuelina canal and Fuel Pool inspection Scope (62707)
The inspectors reviewed two instances where broken glass had fallen into the refueling canal and the spent fuel poo Observations and Findinos Glass from a broken light fixture above the spent fuel pool was found on a fuel assembly stored in the pool. The licenses removed the glass and was unable to determine how the bulb was broken. In a separate instance, a reactor service crane pendant struck a light
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! fixture above the refueling canal and the glass bulb fell into the dry refueling canal and broke while the reactor vessel head was removed. The licensee recovered most of the broken glass from the refueling canal. A video inspection of the reactor vessel did not reveal glass fragments, but the licensee conservatively assumed that the remaining unrecovered glass had entered the vessel. The licensee was in the process of evaluating the potential consequences of the glass being in the vesse A 1996. Quality Assessment Audit of refueling activities identified the potential for clear
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glass to enter into the refueling canal from overhead light fixtures and recommended that bumed out bulbs be replaced with shatterproof bulbs. An MWO was written to replace bumed out bulbs with shatterproof bulbs. At the start of refueling activities, the licensee did not find any bumed out lights near the refueling canal and, therefore, shatterproof bulbs were not installed. Further, fixtures above the spent fuel pool were not replaced with shatterproof bulbs because the audit did not address these light fixture Conclusions
- Two light fixtures were broken during the refueling outage which resulted in broken glass
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entering the refueling canal and the spent fuel pool. This may have been prevented had
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the actions taken to address a 1996 audit finding been more comprehensiv M3 Maintenance Procedures and Documentation
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M Demineralized Water Leakina on Operatina Decav Heat Removal (DHR) Pumo #2 Inspection Scope (62707)
The inspectors reviewed the circumstances surrounding an event where water sprayed onto DHR Pump #2 while the RCS level was being lowered to mid-loop.
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!- Technicians used a temporary manifold to perform a flush of Radiation Elements 1878 A l and B. The manifold was connected to Demineralized Water Supply Valve DW-214 in emergency core cooling system (ECCS) Pump Room #2. The same manifold was then
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used by operators to fill steam generator #2. After the operators completed filling the steam generator, they disconnected the manifold hose that was connected to the steam generator wet layup pump. However, the operators did not close the isolation valve for
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the hose on the manifold. Subsequent to this action, a second flush of the radiation j elements was performeo Dy technicians. The technicians did not notice that a hose connected to the manifold had been disconnected and that its isolation valve was open.
l After the radiation levels were reduced, the technicians retumed to ECCS Pump Room #2
. to restore the system from the flush. During the restoration, the technicians noticed water
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leaking from a pipe chase in the ceiling onto DHR Pump #2. The technicians notified l operations and radiation protection personnel of the problem and PCAQR 98-0613 was written to document the event.'
The source of the water was from the disconnected hose. The technicians missed identifying the open isolation valve during the visual check of the manifold valves because the handle for the valve, indicating its position, was hidden behind another hose. While
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the technicians flushed the radiation elements, water flowed through the open isolation valve out the disconnected hose and onto the floor of Mechanical Penetration Room #2, through the pipe chase, and onto the operating DHR pump.
t The inspectors reviewed Procedures DB-DP-00307, " Station Configuration Control,"
DB-OP-06230, " Steam Generator Secondary Side Fill, Drain and Lay up," and
- - DB-OP-03011. " Radioactive Liquid Batch Release." Procedure DB-OP-00307 allowed
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technicians to operate the temporary manifold valves and did not provide specific guidance on controlling valve positions following maintenance evolutions. Procedure DB-OP-06230 permitted the hose to be disconnected from the wet layup pump. Both Procedures DB-OP-06230 and -03011 provided instructions for controlling the position of j Valve DW-214 but did not provide instructions for controlling the isolation valves on the manifol Appendix B, Criterion V of 10 CFR Part 50 states, in part, that activities affecting quality
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- shall be prescribed by documented instructions of a type appropriate to the
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circumstances and shall be accomplished in accordance with these instruction Contrary to the above, the procedures used for steam generator filling and radiation q element flushing were not appropriate to the circumstances in that instructions for l manipulating the temporary manifold valves used during these activities were not provided. Consequently, the isolation valve for the supply hose to the wet lay up pump j was not closed after the hose was disconnected and, as a result, water from the supply I hose leaked onto the operating DHR pump during RCS draining. This is a violation (VIO 50-346/98005-02(DRP)).
In response to the event, the licensee issued a standing order to control the use of the demineralized water service connections. This included ensuring each manifold had an isolation valve, affixing an informe~ i tag on each valve, ensuring each isolation valve was closed before doing an in-serv 1ce evolution, and stationing a zone operator to ,
monitor the use of the manifold connection. The inspectors concluded that these d corrective actions were acceptable to address the issue with operation and control of the demineralized water temporary manifold connection. However, the licensee did not
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address corrective actions for controlling manifold valves used on other system Conclusion
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The procedures used to fill the steam generator and to flush radiation elements from the temporary manifold were not adequate to ensure all manifold valves were controlled. The i inspectors identified one violatio M8 Miscellaneous Maintenance issues (92902)
M8.1 (Closed) Licensee Event Report 50-346/97-013: Missed Surveillance for the Safety Features Actuation System Channel 2 Containment Radiation Func'ional Unit. Due to licensee identification that a surveillance test was only partially performed before the TS surveillance requirement completion date, the remainder of the test was completed (as provided by TS 4.0.3), interim guidance on documentation of partial test completion was provided to the instrument and Control Supervisors responsible for test performance, and a commitment was made by the licensee to modify the applicable procedure to provide documentation for performance of a partial tes .
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i The inspectors confirmed that the test cover sheet (page 32 of Procedure DB-DP-00013,
" Surveillance and Periodic Test Program,") was modified so that performance of a partial test could be documentad. The inspectors determined that the missed surveillance was a violation of TSs. This non-repetitive, licensee-identified and corrected violation is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50446/9800543(DRP)).
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M8.2 -(Closed) Licensee Event Reoort 50-346/97-014: Pressurizer Pilot-Operated Relief Valve Setpoint Less Than Allowable Value Dua To Degraded Wire Sp, ice. Due to licensee identification of inconsistent operation of instrumentation, a defective wire splice was discovered and repaired. Several days later, the licensee determined that this defective -
splice affected the pressurizer pilot-operated relief valve (PORV) setpoint, but that the l setpoint remained within allowable values.
l l Two and one-half months later, after further review, the licensee identified that the actual L PORV setpoint was less than previously assumed. Since the repair was completed in a ,
timely manner, and since the deviation from the setpoint was small, the identified setpoint condition had minimalimpact on safety. To reduce the probability of a future similar 1 l ~ occurrence, training was provided to appropriate staff, as documented in the licensee's commitment closeout form.
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Each trained individual was required to sign a form stating that he/she had read the training materials. One individual recommended that the applicable procedures should reflect the requirement stated in the training memorandum to minimize the potential for .
future recurrences rather than solely using a training memorandum. After consideration, the licensee changed the procedure to reflect the training memorandum requiremen The inspectors detarmined that the PORV setpoint being less than the allowable value l was a violation of TSs. This non-repetitive, licensee-identified and corrected violation is l being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-346/9800544(DRP))
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E1 Conduct of Engineering E Potential Inadeauste Modification
! Inspection Scope (37551)
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The inspector followed up on the fact that operations personnel were unable to restore 1 J
Service Water Pump #1 on April 25, following modifications to an instantaneous overcurrent rela Observations and Findinas During troubleshooting cctivities, an unexpected inrush of current into a zero-sequence-type cur ent transformer caused a newly modified (initiated by Modification 95-0021) yound fault relay to actuate. The modification of the relay was
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l made to address seismic qualification issues for safe shutdown equipment and consisted of replacing a mechanical high speed trip relay with a solid state high speed relay. In theory, for an ungrounded motor, no current would flow in the current transformer unless a ground fault existed on the load side of the current transformer.
l However, the solid state relay actuated faster than the mechanical relay; therefore, it was more sensitive to transient currents from a pump start. After examining the data, the ,
licensee found that increased sensitivity to transient currents caused the service water I
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pump circuit breaker protective trip. The same type of modification was made to L Component Cooling Water Pump #1 breaker and to the primary breaker for the 480 Volt Essential Bus E-1. Consequently, the licensee declared Service Water Pump #1, Component Cooling Water Pump #1, and the primary breaker for Bus E-1 inoperable until )
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the situation was resolved.
l The inspectors questioned engineering personnel as to whether adequate testing had I
been performed to determine the acceptability of the modification. As a result of the i question, the station generated PCAQR 93-0797. Testing that had been performed on !
the relay included a calibration and a check for proper relay operation after it had been i installed. However, no check of the relay's response relating to actual pump start conditions was performe The inspectors had not completed reviews of pertinent vendor manuals, IEEE standards, j and the safety evaluation for the modification. This is an unresolved item l (50-346/98005-05(DRP)) pending the inspectors' review to determine whether testing, evaluations, the modification package, and the safety evaluation associated with modif,s Jtion of this relay were adequat j Conclusions An unresolved item was opened to determine whether a modification for a relay used in Service Water Pump #1, Component Cooling Water Pump #1 and Essential Bus E-1 i breakers was adequately evaluated, tested and implemente l E2 Engineering Support of Facilities and Equipment E2.1 P_eetina Paint Found on Containment Walls a. Inspection Scope (37551)
The inspectors toured the containment building to review housekeep;ng and general material cendition in the buildin ;
b. Observations and Findinas During routine tours inside the containment building, the inspectors noted localized peeling of paint on the containment and D-ring walls. The licensee also haciidentified blistering, peeling and chipped paint around the refueling canal, on the reactor service crane, and on the missile barrier. The inspectors were concemed that the paint may dislodge and block the emergency sump and restrict or prevent flow to the emergency core cooling system (ECCS) pumps during the recirculation phase of a postulated
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loss-of-coolant-accident. Subsequent!y, the licensee initiated MWOs to correct the condition. This issue is an unresolved item (50-346/98005-06(DRP)) pending the inspectors * review of the licensee's evaluation for past ECCS pump operability and pending a visual inspection during a containment closecut to verify the identified condition has been correcte Conclusions An unresolved item was opened to review the licensee's evaluation of the effect peeling paint had on ECCS pumps and to review the licensee's corrective actions for the conditio E2.2 Q_ust in Mechanical Penetration Rooms and ECCS Rooms (37551)
The inspectors followed up on a condition where a thick layer of dust had accumulated on equipment in the mechanical penetration rooms and ECCS rooms due to sandblasting in the shield building annulus area. Plant engineering and operations personnel conducted ,
inspections in these rooms focusing on the potential degradation to the running decay /
heat removal system pumps ard associated valves. Additionally, design engineering personnel inspected important to-safety conduit, termination boxes, junction boxes, instrumentation, control stations, breakers, motor control centers, and motors. The ,
licensee found that these components were covered or sealed. No degradation to the l electrical system had been noticed as a result of the dust accumulation. The inspectors concluded that engineering personnel conducted a comprehensive inspection and evaluation of the effects of cust on equipment in the affected room E7 Quality Assurance in Ent,ineering Activities E Station Review Board. Manaaement Review Committee. and Quality Assessment Performance (71707)
The inspectors observed meetings of tr.) Station Review Board and Management Review Committee and evaluated the activities of Quality Assessment personnel. Overall, these organizations and individuals were effective in their respective roles. Station Review Board members had appropriately prepared for a May 8 meeting regarding reviews of LER 98-002 and PCAQR 98-0529 involving the April 10 manual trip of the reacto Further, Management Review Committee members appropriately categorized and assigned PCAQRs for completion. Also, Quality Assessment personnel categorized and trended PCAQRs written during the outage to assess the collective significance of similar issues. At least two collective significance PCAORs (reactor cleanliness and crane / rigging operations) were subsequently generated for a comprehensive evaluation of potential common root cause E8 Miscellaneous Engineering issues (92903)
E (Closef) Unresolved item 50-346/96014-02 (DRP): Potential Preconditioning of the Auxiliary Feedwater Flow Control Valve. This URI involved the inspectors' identification that Flow Control Valve AF-6451 for Auxiliary Feedwater Pump #2 had been shut and then opened before stroking the valve to gather stroke time data. The inspectors initially
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considered this valve manipulation as possible preconditioning of the valve and il e licensee issued PCAQR 97-0023 to address the inspectors' concem Subsequently, the licensee issued Procedure Change Requests 97-2376 and -2377 to
' evaluate rearranging the sequence of the procedure steps. After final evaluation, the
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licensee found that the normally deenergized open valves had to be repositioned closed
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to stroke time the valves open. The valves were closed by connecting the voltage
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I; The inspectors determined that timing the valve open without first closing it was unavoidable. Information Notice 97-16, " Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or TS Surveillance Testing,"
stated that certain situations may necessitate repositioning components to perform the '
test and this was not considered intentional preconditioning. The inspectors concluded that the licensee's method for testing Valves AF-6451 and -6452 was consistent with the information Notice 97-16 guidance and was therefore acceptabl E (Closed) Unresolved item 50-346/96006-05 (DRP): Potential Preconditioning of Remotely 4 Operated Valves, This URI involved the inspectors' identification that performance of activities per Procedure DB-SP-03161, " Auxiliary Feedwater Train 2 Level Control,
- Interlock, and Flow Transmitter Test," may involve preconditioning Valve SW-1383 before gathering stroke time data. The procedure directed a verification of to low pressure switch interlock function and stroking of Valve SW-1383 from the Auxiliary Shutdown Pane Consequently, if the procedure was followed in sequence, the valve would have been stroked twice before collecting ASME stroke time dat The licensee initiated PCAQR 96-1318 to address the potential preconditioning issue, in the PCAQR evaluation, the licensee noted that preconditioning was not disallowed by ASME Section XI. However, the licensee concluded that preconditioning the valve was
~ not a good practice because as-found data could be lost, and subsequently changed the .
procedures to prevent preconditionin E (Closed) Inspection Follow-up item 50-346/96010-02 (DRP): Potential Preconditioning of Emergency Diesel Generator (EDG) Air Start Motors. This item involved the inspectors' j identification that during a monthly surveillance test of EDG #1, the licensee used one of two sets of air start motors to bar the EDG. Although the procedure provided stipulations ;
on altemating the air start motor banks monthly to start the EDG, the procedure did not I have a similar restriction on which starting bank to be used for barring. The licensee !
issued PCAQR 96-1463 to address this conce I L
l The inspectors reviewed the EDG #1 and #2 test procedures and noted that the licensee l E changed the procedures to use a manual barring device to bar the EDGs. Since the air i start motors are no longer used to bar the EDG, the ' issue of altemating the air bank motors for barring no longer existed. The inspectors concluded that the procedure !
changes for manual barring of the EDG was acceptabl E (Closed) Unresolved item 50-346/97009-03(DRP): Component Cooling Water Inventory Following a Design Basis Seismic Event. The licensee determined that 20 gpm was the
- acceptance criteria for component cooling water system integrated leakage. During Refueling Outage 11, a component cooling water integrated leak rate test was conducted
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per Procedure DB-PF-04141, "Augmenteci Component Cooling Water Integrated Leakage and Check Valve Leakage Rate Test." The results of the test indicated no leakage.
After further review, the inspectors determined that no regulatory requirements existed for L
integrated leak testing of the valves. Also, no design criteria for component cooling water I integrated leakage existed that would have required validation testing. ASME Section XI did not require leak rate testing of the boundary valves. Because the integrated leak testing demonstrated that the system did not leak, there was no impact to plant safet IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls (71750)
The inspectors routinely observed the control of radiological work activities by personnel at the personnel hatch and during tours inside the containment building. The inspectors observed personnel using appropriate radiation work practices. Additionally, the inspectors noted good health physics support for work activities inside the containment that included: Installing proper lead shielding and radiological postings and barriers as necessary, using video cameras and monitors for remote observation, using containment coordinators and rovers for assistance in containment, and using tele-desimetry to monitor doses to individuals inspecting inside the steam generator V. Manaaement Meetinas X1' Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on May 12,1998. The licensee acknowledged the findings l presented. The inspectc,rs asked the licensee whether any materials examined during the
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inspection should be considered proprietary No proprietary information wss identifie !
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PARTIAL LIST OF PERSONS CONTACTED J. K. Wood, Vice President, Nuclear J. H. Lash, Plant Manager R. E. Donnellon, Director, Engineering and Services L. W. Worley, Director, Nuclear Assurance M. C. Beier, Manager, Quality Assessment C. A. Price, Manager, Business Services R. J, Scott, Manager, Radiation Protection G. R. McIntyre, Manager-Acting, Plant Engineering i
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R. B. Coad, Superintendent, Radiation Protection G. W. Gillespie, Superintendent, Chemistry R. A. Greenwood, Supervisor, Radiation Protection D. H. Lockwood, Supervisor, Regulatory Affairs T. S. Swim, Supervisor, Mechanical / Structural Engineering )
D. L. Miller, Senior Engineer, Regulatory Affairs ;
G. M. Wolf, Engineer, Regulatory Affairs NRC T. J. Kozak, Chief, Reactor Projects Branch 4, R-Ill S. J. Campbell, Senior Resident inspector, Davis-Besse K. S. Zellers, Resident inspector, Davis-Besse I
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i INSPECTION PROCEDURES USED IP 37551: Onsite Engineering ;
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IP 61726: Surveillance Observations H IP 62707: Maintenance Observation (. IP 71707:. Plant Operations Plant Support Activities
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IP 71750:
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IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor i
Facilities IP 92902: Followup - Maintenance IP 92903: Followup - Engineering IP 93702: Prompt Onsita Response to Ev snts at Operating Power Reactors ITEMS OPENED AND CLOSED l
Opened
- 50-346/98005-01(DRP) NCV Violation of Polar Crane Lift Procedure 50-346/98005-02(DRP) VIO Inadequate Control of Temporary Service Manifold Isolation Valves for Supply Hose to Fill the Steam Generato ~ 50/346/98005-03(DRP) NCV Violation due to Missed Surveillance for the Safety Features
- Actuation System Channel 2 Containment Radiation Functional Unit 50-346/98005-04(DRP) NCV Violation of TS due to Pressurizer Pilot-Operated Relief Valve Setpoint Less ThM Allowable Value 50-346/98005-05(DRP) URI Potentialinadequate Modification 50-346/98005-06(DRP) URI Peeling Paint Found in Containment Closed 50-346/98005-01(DRP) NCV Violation of Polar Crane Lift Procedure !
50-346/97-013 LER Missed Surveillance for the Safety Features Actuation _
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System Channel 2 Containment Radiation Functional Unit i 50/346/98005-03(DRP) NCV Violation due to Missed Surveillance for the Safety Features 1 Actuation System Channel 2 Containment Radiation Functional Unit 50-346/97-014 LER Pressurizer Pilot-Operated Relief Valve Setpoint Less Than l Allowable Value due to Degraded Wire Splice l L 50-346/98005-04(DRP) NCV Violation of TS Due to Pressurizer Pilot-Operated Relief {
L Valve Setpoint Less Than TS Allowable Value !
l 50-346/96014-02(DRP) URI Potential Preconditioning of the Auxiliary Feedwater Flow l Control Valve i
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' 50-364/96006-05(DRP) URI - Potential Preconditioning of Remotely Operated Valves
' 50-349/96010-02(DRP) IFl ' Potential Preconditioning of Emergency Diesel Generator Air Start Motor l
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LIST OF ACRONYMS AND INITIALISMS USED-ASME American Society of Mechanical Engineers
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CFR Code of Federal Regulations l
CWRT Clean Waste Receiver Tanks l
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DADS Data Acquisition and Display System DHR Decay Heat Removal ECCS Emergency Core Cooling System EDG Emergency Diesel Generator ICS Integrated Control System IFl Inspection Followup item LER Licensee Event Report MWT Miscellaneous Waste Tank MWO Maintenance Work Order NCV Non-Cited Violation NNI Non-Nuclear Instrumentation NRC Nuclear Regulatory Commission ;
Potential Condition Adverse to Quality Report l
PCAQR PDR Public Document Room PM Preventive Maintenance PORV Pilot-Operated Relief Valve RCDT Reactor Coolant Drain Tank RCS Reactor Coolant System TS Technical Specification URI Unresolved item VIO Violation .
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