IR 05000298/1998008

Insp Rept 50-298/98-08 on 980115-1226.Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20202E337
Person / Time
Site:	Cooper
Issue date:	01/22/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20202E316	List: IR 05000298/1998008 ML20202E312 ML20202E326
References
50-298-98-08, 50-298-98-8, NUDOCS 9902020385
Download: ML20202E337 (15)

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ENCLOSURE 2

• U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50-298 License No.: DPR 46 Report No.: 50-298/98-08 Licensee: Nebraska Public Power District Facility: Cooper Nuclear Station Location: P.O. Box 98 Brownville, Nebtaska  ;

Dates: November 15 through December 26,1998 i i

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inspectors: C. Skinner, Acting Senior Resident inspector

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V. Gaddy, Acting Senior Resident inspector !

N. Garrett, Resident inspector, River Bend Station

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Approved By: C. Marschall, Chief, Project Branch C Division of Reactor Projects '

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ATTACHMENT: Supplemental information

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I EXECUTIVE SUMMARY Cooper Nuclear Station NRC Inspection Report 50-298/98-08 ,

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Operations

The licensee could have avoided two plant shutdowns through better review of the l Technical Specifications Bases for nuclear instrumentation and by properly venting the ]

control rod drive system. Senior management performed good startup briefs, stressing j self-checking, control board awareness, conservative decision making, and effective communications (Section 01.1). ,

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During an observation of maintenance, the inspectors discovered that control room

, operators had authorized work that rendered Residual Heat Removal Loop A ,

l inoperable. Fortunately, plant staff met the Technical Specification 3.5.2 action !

statement requirement until operators restored the inoperable loop. Operators reviewed I and authorized the start of the maintenance without recognizing that it made Loop A inoperable. The operators did not follow operations procedures to log the pumps inoperable when the work was released. This is an example of a violation of Technical Specification 5.4.1 (Section O4.1).

Failure to specify the order in which tags were hung to support integrated leak rate testing resulted in operations personnel lowering reactor vessel water level approximately 1.4 inches before the level decrease was noticed and corrected by control room personnel. Although water was lowered, the core remained sufficiently covere This is an additional example of a violation of Technical Specification 5. (Section 04.2). l

• In June 1997, the licensee reported opening of a torus-to-drywell vacuum breaker as an engineered safety feature actuation. The licensee determined that the inadvertent i opening occurred due to an inadequate procedure. To correct the problem, the licensee l revised the procedure to include a note to warn operators that a differential pressure of 0.25 psig could result in a vacuum breaker opening. Inspectors found that the vacuum breaker surveillance procedure acceptance criteria allowed the vacuum breakers to open between 0.1 and 0.5 psig differential pressure. The inspectors determined that the corrective actions would not have warned the operators of an engineered safety feature actuation at a differential pressure less than 0.25 psig. In addition, the licensee's failure to use the appropriate value from the surveillance acceptance criteria, the Final Safety Analysis Report, or valve design information for the lower pressure threshold of valve ,

opening was a recurrence of the previously identified performance deficiency; therefore, l the inspectors concluded that the licensee did not take reasonable corrective action for the previously identified problem. This is a violation of 10 CFR Part 50, Appendix B, Criterion (Section 08.1). j i

Maintenance l l

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• A reactor recirculation pump seal failed due to improper installation of the thrust collar.

l The work instructions contained two steps that would have identified the thrust collar not being installed properly. Workers marked one step N/A and did not perform the other l l

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step correctly. The licensee identified that the workers did not understand the technica!  ;

requirements and took corrective action. This is an additional example of a violation of -

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Technical Specification 5.4.1 (Section M1.2).

• A reactor vessel low water level alarm signal was received while restoring instruments to service after backfilling a common reference leg. The maintenance technicians were only to restore the vessel nozzle range instruments, but the procedure did not separate the vessel nozzle range and flood-up instruments. When the flood-up instruments were restored, air displaced the water column, causing the indicated low water level. This is a j noncited violation of to CFR Par 150, Appendix B, Criterion V (Section M8.1).

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• The licensee discovered that 23 instrument surveillances supporting emergency core cooling systems had exceeded their Technical Specification functional check requirement. No instruments were subsequently identified as being out of toleranc l This is a noncited violation of Technical Specification 4.2.B (Section M8.2).

l * The licensee discovered that the high flux trip setpoint was not being tested as require l The flow biased trip setpoint was being tested instead, due to a lack of understanding of '

the difference between the two trip setpoints. This is a noncited violation of Technical l Specification 4.1.A (Section M8.3).

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Enaineerina

• The licensee discovered that the pressure rating for downstream nonsafety-related pressure regulators which provided motive force for primary containment valves on the l reactor coolant sample line would be exceeded. This would result in the failure of the primary containment valves being able to close. This is a noncited violation of 10 CFR Part 50, Appendix B, Criterion ill, Design Control (Section E8.1).

Plant Suocort

• A fire watch failed to protect all cornbustible material and openings in walls and floors within 35 feet of the hot work area with noncombustible blankets or shields to prevent I passage of sparks to adjacent areas as required by procedures. This is an additional example of a violation of Technical Specification 5.4.1 (Section F1.1).

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Report Details Summary of Plant Status At the beginning of the inspection period, the plant was shut down for Refueling Outage 1 Plant startup was commenced on December 15,1998, but was terminated when overlap l between the source range and intermediate range monitors was not sufficient. Plant startup was attempted again on December 16; however, a control rod failed to function and all control rods were inserted. On December 17, the reactor went critical and Mode 1 was entered on i December 18. On December 20, the main generator was synchronized to the grid and the <

plant reached 100 percent power on December 23. The plant remained at 100 percent for the l rest of the inspection perio . Operations I 01 Conduct of Operations l

0 Plant Startuo Inspection Scope (71707)

The inspectors observed the licensee's attempts to make the reactor critica Discussions were held with operations, reactor engineering, and plant managemen j l Observations and Findinas On December 15,1998, with the plant in Mode 2, the inspectors observed the approach to criticality. Plant startup was conducted in accordance with Procedure 2.1.1, "Startup Procedure." Prior to the startup, plant managers and the shift supervisor conducted l briefs. The briefs covered precautions, industry experience, and management l expectations. Senior management also stressed self-checking, control board awareness, conservative decision making, maintaining a broad focus, and effective communication Technical Specification Surveillance 3.3.1.1.5 required that source range monitors and 1 intermediate range monitors have overlap. The Technical Specification 3.3.1.1.5 Bases ;

for this surveillance requirement states that overlap exists between source range I monitors and intermediate range monitors when, with the source range monitors fully inserted, the intermediate range monitors are above midscale of Range 1 before the .

l source range monitors have reached the upscale rod block. During the startup, not all of the intermediate range monitors reached midscale of Range 1 before the source range monitors reached the upscale rod bicck. As a result, operators shut the plant down. The licensee changed the Technical Specification Bases of Surveillance Requirement 3.3.1.1.5 to state that, prior to withdrawing the source range monitors from the fully inserted position, all operable intermediate range monitor channels shall be on scal On December 16,1998, operators placed the mode switch in startup and began withdrawing control rods to achieve criticality. They successfully withdrew the first two

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2-control rods but the third control rod,30-19, could not be withdrawn beyond position 0 The operators entered Abnormal Procedure 2.4.1.1. " Stuck or inoperable Control Rod,"

and reinserted all control rods. Operators and engineering personnel attributed the control rod problem to irnproperly venting the control rod drive system. After correcting the problem, operators exited Abnormal Procedure 2.4.1.1 and resumed the approach to criticality. The reactor reached criticality at 1:15 a.m. on December 1 Conclusion The licensee could have avoided two plant shutdowns through better review of the Technical Specifications Bases for nuclear instrumentation and by properly venting the control rod drive system. Senior management performed good startup briefs, stressing self-checking, control board awareness, conservative decision making, and effective communication Operator Knowledge and Performance 04.1 Failure to Recoanize Maintenance Effect on Operable Eauipment Inspection Scope (71707)

The inspector held discussions with operations and engineering on Technical Specification 3.5.2, which requires two emergency core cooling systems to be operable while in Modes 4 or Observations and Findinos On December 2,1998, the inspectors identified that both methods for Residual Heat Removal Loop A room cooling were inoperable. The opening for natural air circulation required for room cooling was 75 percent covered and maintenance on the room cooler secured the reactor equipment cooling system flo At the time of this maintenance, the control room crew was taking credit for one of the ;

residual heat removal pumps in Loop A as the second emergency core cooling system, required by Technical Specification 3.5.2. Residual heat removal pumps in Loop A have two ways to be cooled to ensure pump operability, the room cooler and natural air circulation. Maintenance on the room cooler rendered the room cooler inoperabl Technicians covered 75 percent of an opening which rendered the natural air circulation inoperable. Technical Specification 3.5.2 required that, if two pumps were not operable, the licensee must initiate actions to suspend operations with a potential for draining the reactor vessel. The licensee had already initiated these requirements based on a separate Technical Specifications action statemen The inspectors identified that the maintenance was released by operations without recognizing the impact the work would have on Residual Heat Removal Loop A. Since the control room crew did not know that the plant was not meeting the Technical Specifications limiting condition for operation, the operators did not tog the pumps

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-3-inoperable as required by Step 8.8 of Procedure 2.0.2," Operations Logs and Reports,"

Revision 44. The failure to perform Step 8.8 of Procedure 2.0.2 is an example of a violation of Technical Specification 5.4.1 (50-298/98008-01). This violation is being cited due to the identification of several examples of failure to follow procedures in this report and in NRC Inspection Report 50-298/98-0 Conclusion During an observation of maintenance, the inspectors discovered that control room operators had authorized work that rendered Residual Heat Removal Loop A inoperable. Fortunately, plant staff met the Technical Specification 3.5.2 action statement requirement until operators restored the inoperable loop. Operators reviewed and authorized the start of the maintenance without recognizing that it made Loop A inoperable. The operators did not follow operation procedures to log the pumps inoperable when the work was released. This is an example of a violation of Technical Specification 5. .2 Inadvertent Lowerina of Reactor Vessel Level Inspection Scope (71707)

The inspectors reviewed the circumstances surrounding an inadvertent lowering of reactor vessel water level on December 5,199 _Ooservations and Findinas On December 5,1998, control room operators noted that reactor vessel water level had lowered approximately 1.4 inches, as indicated by a control room level indicato Operators performed an evaluation and determined that reactor vessel water level was being lowered beca'ise Core Spray B Manual isolation Valve CS-V-14A was open, allowing level to drain from the reactor to the core spray syste i While performing a core spray system tagout, operators had opened Valves CS-172 and -17'3 (Inboard Test Connection isolation and Inboard Test Connection Supply)

before they closed Valve CS-V-14A. With these valves open, water was diverted from the reactor vessel to the core spray system. Although water level had lowered approximately 1.4 inches, it had remained in the normal range for water leve The inspectors determined that the tagging order did not specify the order to hang the l

tags. The inspectors reviewed Administrative Procedure 0.9," Tagging Orders,* and '

noted that Step 5.1 required tags to be hung and removed in a specified order. Failing to specify the proper order for hanging tags is an additional example of failure to follow procedures (50-298/98008-01).

As corrective action, operators closed Valve CS-V-14A and reduced reactor water cleanup blowdown flow to restore water level. In addition, plant staff initiated a Problem ,

Identification Report and discussed the problem with all of the individuals involve l Operators established a specified sequence for hanging the remainder of the tags for l

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-4-the core spray system and assigned a specific individual the responsibility for coordinating the remainder of integrated leak rate test activitie Conclusions Plant staff failed to specify the correct order to hang tags for a tagout. As a result, reactor vessel water level dropped approximately 1.4 inches before operators noticed l the level decrease and corrected the condition. The reduction in water level did not I impact core coverage. This is an additional example of a violation of Technical Specification 5.4.1.

l 08 Miscellaneous Operations issues (92901)

l 0 (Closed) Licensee Event Report 50-298/97-007: The licensee reported opening of a torus-to-drywell vacuum breaker as an engineer safety feature actuation. The vacuum i breaker opened as a result of the differential pressure caused by inerting the drywell l (replacing the air with nitrogen). The licensee determined that the inadvertent opening occurred due to an inadequate procedur To correct the problem, the licensee added a note to Procedure 2.2.60, " Primary Containment Cooling and Nitrogen inerting System," to warn operators that a differential pressure of 0.25 psig may result in a vacuum breaker opening. Additional steps limited the differential pressure between the torus and drywell to less than 0.15 psig. No unplanned opening of torus-to-drywell vacuum breakers has occurred as a result of inerting the drywell during the last three plant startup Inspectors found that the vacuum breaker surveillance Procedure 6.PC.308,

! " Suppression Chamber-to-Drywell Vacuum Breaker Calibration and Functional Test,"

Revision 4, acceptance criteria allowed the vacuum breakers to open between 0.1 and

0.5 psig differential pressure. The inspectors determined that the corrective actions l would not have warned the operators of an engineered sa'fety feature actuation at a differential pressure less than 0.25 psig. The additional steps did not limit the differential pressure to less than 0.1 psig, the lower limit for acceptable valve operatio Since the licensee did not use the appropriate value from the surveillance acceptance criteria, the Final Safety Analysis Report, or valve design information for the lower pressure threshold of valvo opening, the inspectors concluded that the licensee did not take reasonable corrective action for the previously identified problem. The inspectors noted that, although no additional unplanned valve opening had occurred, the corrective actions did not preclude valve openings for drywell pressures between 0.1 psig (the lower limit in the surveillance acceptance criteria) and 0.15 psig (the upper drywell pressure limit during inerting, imposed by procedure).

Previous NRC inspection reports noted that the Cooper staff has improved their l effectiveness in documenting conditions adverse to quality in problem identification reports. The inspection reports have also noted; however, that the Cooper staff has  ;

l demonstrated significant lack of effectiveness in correcting the identified problems. In .

particular, NRC Inspection Report 50-298/97-07 documented several failures to identify, j

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correct, determine the cause, or prevent recurrence of conditions adverse to qualit !

That report also discussed examples of lack of consideration of design information in addressing questions of equipment operabilit )

i The failure to implement effective corrective actions to prevent the inadvertent opening of torus-to-drywell vacuum breakers is a violation of 10 CFR Part 50, Appendix B, Criterion XVI (50-298/98008-02). This violation is being cited because the licensee did ,

not take reasonable corrective action to address the problem identified by a previous l l

violation (50-298/98007-07).

l 11. Maintenance M1 Conduct of Maintenance l l

M1.1 General Comments Inspection Scope (62707)

The inspectors reviewed procedures and work packages and held discussions with maintenance, operations, and managemen I Observations and Findinas The inspectors reviewed and/or observed the following procedures:

Procedure 6.1DG.101, " Diesel Generator 31-Day Operability Test" Procedure 6.PC.532, " Operations Activities for the Primary Containment integrated Leak Rate Test" Procedure 6. MISC.502, "ASME Class i System Leakage Test" Procedure 6.PCIS.301, "PCIS Group 2, ?, and 6 isolation Logic System Functional Test and Reactor Building Vent Monitor Fur'ctional Test" Procedure 7.2.5.1," Reactor Recirculation Pump Seal Cartridge Removal and Installation" Maintenance Work Request 98-3542, " Scram Outlet Valve Maintenance (CRD 18-02)"

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l Maintenance Work Request 98-4003," Replacement of Reactor Equipment Cooling System Piping and Valves" Conclusions The inspectors did not identify any concerns with the maintenance activities listed above, except those discussed in other sections of the report.

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-6-M1.2 Reactor Recirculation Pumo B Seal Failure l

Inspection Scope (62707 and 93702)

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a-The inspectors reviewed the circumstances associated with the failure of the Reactor

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Recirculation Pump B seal. Discussions were held with operations, maintenance, and ;

engineering personne l i Observations and Findines On December 13,1998, during the performance of Surveillance 6. MISC.502,"ASME Class 1 System leakage Test," the licensee received a high sealleakage flow alarm on l the Reactor Recirculation Pump B sea !

Maintenance personnel disassembled the seal and discovered an improperly installed thrust collar. When workers replaced the seal earlier in the refueling outage, they had ,

not performed Step 8.2.38 of Procedure 7.2.5.1," Reactor Recirculation Pump Seal :

Cartridge Removal and Installation." The workers had marked Step 8.2.38, used to check the thrust button clearance as N/A. The discrepancy sheet documented that  ;

workers did not perform this step because the vendor had indicated it was not needed j

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during normal seal maintenance. This step identified acceptance criteria for the thrust button clearance of between 0.002 and 0.004 inch. The plant staff implemented the vendor recommendations without evaluating the effect on the seal. When workers  :

disassembled the seal, they found a thrust button clearance of 0.24 inch. The licensee concluded that, if workers had performed Step 8.2.38, they would have discovered that j they had improperly installed the thrust colla !

in addition, Step 8.2.39 directed maintenance personnel to install the thrust disc and spacer and ensure alignment of the match marks on the flanges. Workers signed the step as complete and quality control staff witnessed the completion. The licensee concluded that the workers had not performed the step correctly, since correct alignment of the match marks required thrust button clearance within the acceptance criteria of Step 8.2.3 The licensee concluded that the workers who had replaced the seal did not fully understand the technical requirem6nts involved with replacing the seal. As a result, the-licensee took corrective action for the mechanics and quality control inspector involved in the seal replacemen Failing to perform steps in Procedure 7.2.5.1 that would have insured proper installation of the thrust collar is an additional example of a violation of Technical Specification 5. (50-298/98008-01). j Conclusions A reactor recirculation pump seal failed due to improper installation of the thrust colla The work instructions contained two steps that would have identified the thrust collar not being installed properly. Workers marked one step N/A and did not perform the other

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-7-step correctly. The licensee identified that the workers did not understand the technical requirements and took corrective action. This is an additional example of a violation of Technical Specification 5. M Maintenance and Material Condition of Facilities and Equipment M2.1 Torus and Drvwell Insoection Inspection Scope (62707. 71750)'

The inspectors performed inspections of the torus and drywell prior to final closeou Observations and Findinas On December 10,1998, the inspectors accompanied the drywell coordinator on an inspection of the pressure suppression pool (torus). During the inspection only small pieces of debris were noted. This material was immediately removed from the are The inspectors noted that the paint coating on the housing of the 12 torus-to-drywell vacuum breakers was flaking in certain areas. In response to the inspector's observation, the licensee initiated Problem Identification Report 3-20647. Resolution of the problem identification report required that workers remove the flaking paint from the vacuum breaker On December 13 and 14, the inspectors performed a drywell walkdown. The inspectors l noted that, generally, the drywell was free of major debris. The small pieces of debris that were identified were immediately removed. The inspectors noticed a leak of several drops per minute from the insulation surrounding Valve FW-CV-16 (Feedwater Check Valve). In response, the licensee stopped the leak by removing the insulation and tightening the body-to-bonnet bolt Conclusions Overall, inspectors considered the pressure suppression pool and drywell very clean and in good material condition. Minor issues the inspectors identified were immediately correcte M8 Miscellaneous Maintenance issues (92902,92700)

M8.1 (Closed) Licensee Event Report 50-298/97-004-00 and -01: A reactor vessellow-water level occurred as technicians restored instruments to service. The low-level signal resulted in a full scram and Group 2 (primary containment), Group 3 (reactor water cleanup), and Group 6 (Secondary containment and standby gas treatment) isolation The Group 2 isolation resulted in a loss of shutdown coolin While restoring the vessel nozzle range instrument to service, technicians erroneously restored the flood-up instrumentation to service. This allowed air to displace the water column in the instrument variable leg, causing an indicated low level. The licensee

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• l 8-I determined that the technicians backfilling the reference leg did not stop to question circumstances that differed from what they expected. An inadequate prejob brief and j inaccurate procedures that did not effectively address human factors contributed to the i cause. As corrective actions, the licensee discussed the circumstances with instrument l and control personnel and revised reference leg backfill procedures to require a detailed '

prejob brief. In addition, they revised operations and maintenance procedures directly related to this event to enhance human factor The inspectors reviewed the licensee's evaluation and concluded that Procedure 14.4.8, i

" Instrument Sensing Line Reference Leg Condensing Pot 1 Backfill," was not I appropriate to the circumstances. Procedure 14.4.8 directed performers to remove and later restore both vessel nozzle range and flood-up instruments. The inspectors I observed that the corrective actions also addressed the procedure inadequacie !

The failure to ensure that Procedure 14.4.8 restored the required instrumentation is a violation of 10 CFR Part 50, Appendix B, Criterion V, " Procedures." This nonrepetitive, .

licensee-identified, and corrected violation is being treated as a noncited violation, !

consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-298/98008-03). I M8.2 (Closed) Licensee Event Report 50-298/97-006: The licensee discovered that they had not performed 23 emergency core cooling system instrument functional checks within their required surveillance intervals. The root cause evaluation concluded that l personnel error in interpreting Technical Specification requirements for scheduling surveillances caused the errors. To prevent recurrence the licensee developed: (1) a matrix that identifies reactor modes of operation and the Technical Specifications '

operability requirements for each mode; (2) a desk guide that lists expectations and duties of the surveillance coordinator; and (3) a computer program, Premo Pas Surveillance developed to ensure compliance with Technical Specification surveillance requirements. The inspectors considered the corrective actions appropriat The failure to perform surveillances within the required intervalis a violation of Technical Specification 4.2.B. This nonrepetitive, licensee-identified, and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-298/98008-04).

M8.3 (Closed) Licensee Event Report 50-298/98-005-00 and -01: Plant staff did not functionally test the average power range monitor system high flux reactor protection system trip setpoint as required by Technical Specification 4. Technical Specification 4.1.A required testing of flow biased trip setpoint once per month and the high flux trip setpoint once per week. The licensee determined that the flow biased trip setpoint was being tested instead of the high flux trip setpoint. The licensee identified that the root cause was a lack of understanding of the commonality between the flow biased and fixed bias circuitry, which led to the deficient surveillance procedur The licensee immediately completed the required surveillance with acceptable result In addition, they revised the appropriate setpoint procedures to ensure testing of the high flux trip setpoint once per week. Inspectors noted that technicians had routinely

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-9-performed successful calibrations of the high flux trip setpoint, indicating that the average power range monitors had remained operable despite the missed surveillance The failure to test the high flux trip set point is a violation of Technical Specification 4.1.A. This nonrepetitive, licensee-identified, and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-298/98008-05).

Ill. Enaineerina E8 Miscellaneous Engineering issues (92903)

E8.1 (Closed) Licensee Event Report 50-298/96-015: The licensee identified an inadequate pressure rating for the nonsafety-related air pressure regulators that provide the motive force for valves on the reactor coolant sample. The excess pressure could expose the two redundant air-operated containment isolation valves to more than their rated pressure. This could result in a failure that prevents the valves from closin Inspectors reviewed this item in NRC Inspection Report 50-298/97-08, which documented that the nonsafety-related regulators were replaced with safety-related regulators. Inspectors did not close this item because they questioned why plant staff had removed a safety-related relief valve in conjunction with the corrective actions for the problem identified in the licensee event report. Plant staff removed the relief vise because it was no longer required. The inspectors learned that safety-related Pressure Relief Valve PC-PRV-PCV632 adequately performed the pressure relief function of the valve previously removed. The inspectors did not identify any additional concern The failure to ensure adequacy of the design for the containment isolation valves is a violation of 10 CFR Part 50, Appendix B, Criterion lil, Design Control. This nonrepetitive, licensee-identified, and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-298/98008-06).

IV. Plant Support F1 Control of Fire Protection Activities F Failure to Follow Fire Protection Procedures Inspection Scope (71750)

The inspectors observed hot work in the reactor building. Discussions of the inspectors'

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. Observations and Findinas On December 2,1998, the inspectors observed technicians removing reactor equipment cooling system piping and valves in accordance with Maintenance Work Request 98-400 Procedure 0.39, ' Fire Watch," Revision 18, Step 8.1.4.2, stated, in part, that combustible material below and for a radius of 35 feet must be protected from hot wor Step 8.1.4.3 required that workers protect all openings in walls and floors within a 35-foot radius of the hot work area with noncombustible blankets or shields * prevent passage of sparks through the openings. The inspectors identified that the fire watch did not comply with these steps. Unprotected rags, tape, and plastic bags remained within the 35-foot radius, and part of a grating and the opening around a pipe passing through the floor were not covered potentially allowing sparks to pass throug On December 8, the licensee identified another example of an inadequately covered opening within a 35-foot radius of hot work. The licensee wrote problem identification reports,3-53687 and 3-53686, to identify these issues and place them in the licensee's corrective action progra The failure of the fire watch to protect all combustible material and openings within a 35-foot radius as required by Procedure 0.39 is an additional example of a violation of

. Technical Specification 5.4.1 (50-298/98008-01). Conclusions

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A fire watch failed to protect all combustible material and openings in walls and floors within 35 feet of the hot work area with noncombustible blankets or shields to prevent passage of sparks to adjacent areas as required by procedures. This is an additional example of a violation of Technical Specification 5. X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the exit meeting on December 31,1998. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie _ _

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SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED

, Licensee M. Bergmeier, Operations Manager D. Buman, Acting Plant Engineering Department Manager T. Chard, Radiological Manager L. Dewhirst, Licensing Engineer P. Donahue, Engineering Support Department Manager L. Dugger, Design Engineering Assistant Manager-C. Fidler, Assistant Maintenance Manager

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B. Houston, Licensing Manager J. McMahan, Acting Work Control Manrger L. Newman,' Assistant Plant Manager

'M. Peckham, Plant Manager _

' J. Peters, Licensing Secretary i B. Rash, Senior Engineering Manager A. Shiever, Operations Manager INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observation IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92700: LER - Onsite Review IP 92901: Followup - Operations IP 92903: Followup - Engineering IP 93702: Onsite Response ITEMS OPENED. OPENED AND CLOSED. AND CLOSED j i

Opened l 50-298/98008-01 VIO Four examples of failure to follow procedures (Sections 04.1, 04.2, M1.2, and F1.1).

50-298/98008-02 VIO Actions to correct an inadequate procedure which resulted in the opening of a torus-to-drywell vacuum breaker would not have prevented recurrence under all conditions (Section O8.1). l Opened and Closed 50-298/98008-03 NCV Reactor trip signal, engineered safety feature actuation, and loss of shutdown cooling during maintenance (Section M8.1).

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-2-50-298/98008-04 NCV Failure to perform emergency core cooling system suiveillances within required surveillance interval (Section M8.2).

50-298/98008-05 NCV Technical Specification violation due to inadequate testing of average power range monitoring system (Section M8.3).

50-298/98008-06 NCV Potential for single active failure in reactor recirculation system (Section E8.1).

Closed 50-298/97-007 LER Opening of a torus-to-drywell vacuum breaker (Section 08.1).

50-298/97-004-00 50-298/97-004-01 LER Reactor trip signal, engineered safety feature actuation, and loss of shutdown cooling during maintenance (Section M8.1).

50-298/97-006 LER Failure to perform emergency core cooling system surveillances within required surveillance interval (Section M8.2).

50-298/98-005-00 50-298/98005 -01 LER Technical Specification violation due to inadequate testing of average power range monitoring system (Section M8.3).

50-298/96-015 LER Potential for single active failure in reactor recirculation system (Section E8.1).

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