IR 05000263/1996007
| ML20128M506 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/07/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20128M461 | List: |
| References | |
| 50-263-96-07, 50-263-96-7, NUDOCS 9610160107 | |
| Download: ML20128M506 (19) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No.:
50-263
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License No.:
DPR-22 Report No:
50-263/96007 Licensee:
Northern States Power Company Facility:
Monticello Nuclear Generating Station Location:
414 Nicollet Mall Minneapolis, MN,55401 Dates:
July 21 - August 28, 1996 Inspectors:
A. M. Stone, Senior Resident Inspector J. Lara. Resident Inspector Approved by:
M. Jordan, Chief, Projects Branch 7 l
Division of Reactor Projects
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i 9610160107 961007 PDR ADOCK 05000263
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EXECUTIVE SupWIARY
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Monticello Nuclear Generating Station, Unit 1 l
NRC Inspection Report 50-263/96007
This !ntegrated inspection included aspects of licensee operations, f
engineering, maintenance, and plant support performed by the resident i
inspectors.
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l Operations i
Operators responded appropriately to a reactor water cleanup system i
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instrument line leak (Section 01.2).
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Operations personnel used a superseded surveillance test due to
personnel error. However, the correct revision of the surveillance was
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identical to the one performed (Section 03.1).
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The inspectors verified the core spray system was operated and l
maintained according to the Updated Final Safety Analysis Report and Technical Specifications requirements (Section 02.2).
Maintenance The maintenance work activities were performed well. The completed j
surveillance tests were verified to have met the test acceptance criteria (Section M1.1).
The licensee adequately implemented controls on overtime. Discrepancies
noted were also identified by quality services personnel and documented in an observation report (Section M7.1).
Enaineerina The inspectors concluded that the licensee's August 16 evaluation for
the reactor water cleanup high energy line break discrepancy was l
acceptable. The licensee's August 22 evaluation is still under review (Section E2.1).
The system engineer assisted operations by providing a sound operability
i evaluation for the #12 emergency service water pump (Section E2.2).
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Plant Suncort The licensee responded aggressively to an industry concern regarding
confidentiality of operator examinations (Section S6.1).
The fire brigade and control room personnel responded promptly and
l appropriately during a fire drill (Section F4.1).
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Report Details l
Summary of Plant Status The plant operated at or near full power for the entire inspection period.
Short term power reductions were conducted during the inspection period for
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control rod adjustments and surveillance testing.
I. Doerations
Conduct of Operations
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01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of plant operations.
In general, the conduct of operations was conservative; specific events and noteworthy observations are detailed in the sections below.
In particular, the inspectors noticed good attentiveness to control room panels and indications. Operators were knowledgeable of equipment status and current operational issues.
The inspectors observed plant operators during daily rounds and a hot area inspection of the main steam vault.
The operators were knowledgeable of plant conditions and initiated actions to correct discrepant equipment conditions.
The operators performance during surveillance activities, problem identification and response to abnormal plant conditions was determined to be acceptable.
01.2 Ocerator Response to Reactor Water Cleanuo (RWCU) Instrument Line Leak On August 25, 1996, plant personnel identified water on the floor near the hydraulic control units.
Plant operators determined the source of the water was from a small leak on a RWCU local pressure indicator instrument line inside of the RWCU room.
The inspectors interviewed i
control room operators to assess their response to the leak. This event i
was of particular interest due to a current RWCU high energy line break (HELB) issue as discussed in section E2.1.
The leak did not cause conditions in the room to reach the room temperature or area radiation alarm setpoints. The increased room temperature and radiation level was minimal. Therefore, control room a
i operators were unaware of the small leak until notified by plant personnel. The control room operators then monitored room temperature and area radiation monitors according to abnormal procedures. The plant operators isolated the leaking instrument line by closing the associated root valve. The inspectors determined that operator response demonstrated awareness of the current RWCU HELB issue.
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Operational Status of Facilities and Equipment
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02.1 Goerational Reviews a.
Insoection Scone (71707)
The inspectors reviewed licensee response to various operations issues and system walkdowns.
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b.
Observations and Findinas l
Control Valve Limit Linkaae i
On August 12, an operator performing a routine turbine floor inspection
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identified a broken roll pin used to connect the turbine control valve limit linkage. The linkage had separated and was hanging off to one side. Maintenance personnel removed the broken roll pin and re-connected the linkage with a new pin.
No adjustments were made to the linkage during the re-installation of the pin thereby maintaining the previous linkage setting.
The licensee performed a preliminary review and determined the condition was not reportable per 10 CFR 50.72. The licensee initiated a condition report to document the root cause and corrective actions. The inspectors independently reviewed the licensee's Updated Final Safety Analysis Report (UFSAR) Chapter 14 and Core Operating Limit Report and determined that the licensee's preliminary reportability evaluation was acceptable.
Enaineered Safety feature System Walkdowns The inspectors walked down selected portions of the core spray, emergency diesel gener:ator emergency service water, and standby gas treatment systems. The inspectors did not identify any conditions that impacted equipment operability.
c.
Conclusions Good operator awareness of plant material condition was evident through identification af the disconnected limit linkage.
02.2 Core Soray $vstem a.
Insoection Scone (71707 and 61726)
During this inspe.ction period, the inspectors reviewed the core spray (CS) system. The purpose of this review was to verify various operational and design features including the following:
operation of the system was according to applicable technical
specifications (TS) and UFSAR requirements; 4 !
system alignment according to plant requirements;
l surveillance procedures met the TS requirements for surveillance
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testing; I
system logic bypasses and permissives for system operation were as
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described in the UFSAR; calibration procedures provided correct acceptance criteria and
correctly specified whether the CS system was rendered inoperable; and plant modifications on the CS system contained appropriate 10 CFR a
50.59 review.
The inspectors reviewed the following documents:
UFSAR 6.2.2 Reactor Core Spray Cooling System 7.1.1.2 Adequacy of the Monticello Emergency Core Cooling Systems (ECCS) Designs 14.7.2.3.1.1 Core Spray System
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3.5.A/4.5.A Core and Containment Spray / Cooling Systems 3.13.H/4.13.H Alternate Shutdown System
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Table 3.2.2 Instrumentation That Initiates ECCS Systems Table 4.2.1 Miniram Test and Calibration Frequency for Core Cooling Rod Block and Isolation Instrumentation Surveillance 0034 ECCS Valve Permissive Sensor 0037 APRS - Low Pressure Core Cooling Pumps Discharge Pressure Interlock Instruments Test and Calibration 0098 Core Spray Header Differential Pressure Test and Calibration Procedure 0113-2 Automatic Depressurization System 20 Minute Timer Test 0419-2 ASDS Core Spray and 14 Emergency SW System Functional Test 2154-11 Core Spray System Prestart Valve Checklist 7120 Core Spray System Instrument Maintenance Procedure 0255-03-IA-1 Core Spray System Tests 0255-03-IIB Core Spray System Pressure Tests Operations Manual B.3.1, Core Spray Cooling System Design Bases Document B.3.1, Core Spray System b.
Observations and Findinas The two independent CS systems were designed to provide a source of injection into the reactor vessel following a design basis accident.
l Each of the two CS loops consisted of a pump, valves, and associated i
piping and controls.
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The inspectors verified that the system operation was as described in the licensee's UFSAR. Additionally, the inspectors verified the following design features:
initiation logic included 20 minute timer on reactor low-low water
level; independence of power sources;
auto-closure logic upon CS initiation signal for test bypass
valve; low reactor pressure interlock and logic for isolation valves and
installation of diverse type pressure switches; and CS valve bypass switch and logic for containment isolation.
- The inspectors verified that TS surveillance requirements were properly translated into surveillance procedures. The inspectors also raviewed flow diagrams, elementary drawings, and instrument calibration records for accuracy.
The inspectors reviewed modification 95Q215, " Low Pressure Coolant i
Injection loop Select and Reactor Pressure Instrument Improvement."
This nodification replaced a CS system pressure switch that provided a
low r u ctor pressure valve permissive signal. The licensee appropriately addressed UFSAR impact and completed an acceptable 10 CFR
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50.59 review. The inspectors independently performed a field inspection
of the CS system including CS ccmponents, CS controls and indications in the main control room, and alternate shutdown panel controls. The inspectors also reviewed temporary jumpers and hold cards to identify any potential. operability concerns.
No deficiencies were identified
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during this review.
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Conclusions
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The inspectors concluded that the CS system was being operated and
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tested according to TS requirements and as described in the UFSAR.
Surveillance procedures appropriately incorporated TS surveillance requirements. System instruments were being appropriately calibrated according to TS requirements.
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Operations Procedures cad Documentation 03.1 Incorrect Procedure Revision Performed a.
Inspection Scone (71707)
The inspectors reviewed surveillance 0085, " Standby Liquid Control System Operability Test," revisions 21 and 22.
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Observations and Findinas In the evening of August 5, operations personnel used revision 21 of surveillance test 0085. The inspectors, while observing the
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surveillance, identified that the operators should have performed the surveillance using revision 22.
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The inspectors determined the following:
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Revision 21 contained a temporary procedure change that eliminated
the requirements for locks on some valves.
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Revision 22 incorporated these temporary procedure changes and was
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implemented on August 5, 1996.
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21 from the shelf and updated the computer to indicate receipt of
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revision 22.
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The operations support personnel retrieved revision 21 of this
procedure about a week earlier. The shift supervisor was expected
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to verify proper revision before implementing.
The shift l
supervisor did not identify that revision 21 was superseded.
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Conclusion a
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The inspectors concluded that the two tests were identical. Therefore, i
performing the wrong revision did not impact system operability. This j.
failure to perform the correct revision constituted a violation of minor significance and is being treated as a Non-Cited Violation consistent
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with Section IV of the NRC Enforcement Policy (50-263/96007-01).
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Miscellaneous Operations Issues 08.1 (Closed) Unresolved Item (50-263/95006-01): A quality control inspector noted that operations personnel did not have an approved procedure for resetting the recirculation pump motor generator (MG) set scoop tube.
The operators reset the scoop tube positioner using informal instructions provided by instrument and control personnel.
The inspectors reviewed the licensee's corrective actions including a new procedure, 7091, " Reset of Scoop Tube with Recirc Pump Speed Above Minimum," revision 0.
This procedure was approved by the operations committee in September 1995. This procedure was deleted after the 1996 refueling outage due to modifications to the MG set control system. The inspcctors verified that procedures were implemented for this new control system.
Technical specification (TS) 6.5.A.1 required procedures for operations of all components involving nuclear safety and TS 6.2.B.4 required the operations committee to periodically review such procedures. This failure to use an approved procedure for resetting the recirculation pump MG set constituted a violation of minor significance and is considered a Non-Cited Violation consistent with Section IV of the NRC Enforcement Policy (50-263/96007-02).
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II. Maintenance M1 Conduct of Maintenance M1.1 General Comments a.
Insoection Scone (62703)
The inspectors observed all or portions of the following work orders (W0) and surveillance activities:
WO 9601980 Install Reinforcing Steel to Reactor Building Roof and Columns
WO 9602160 Perform Maintenance on K-9A Compressor Motor
WO 9602242 B-RBV-WRGM V-EF-20 probe out of as found
WO 9602243 Replace B Recirc Relays In CRV-EFT System
WO 9602244 Replace B High Rad Relays In CRV-EFT System
0040a Average Power Range Monitor Monthly / Rod Block Functional Test
0081 Control Rod Drive SCRAM Insertion Time Test
0085 Standby Liquid Control System Operability Test
0187-2 12 Emergency Diesel Generator /12 Emergency Service Water System Tests
0363-2 RBV Wide Range Gas Monitors Process and Sample Flow Instrument Calibration Procedure 7110 Residual Heat Removal System Instrument Maintenance
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Observations and Findinas The inspectors found the work performed under these activities to be professional and thorough. All work observed was performed with the work package present and in active use. Technicians were exper:enced and knowledgeable of their assigned tasks. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control personnel were present whenever required by procedure.
When applicable, appropriate radiation control measures were in place.
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Conclusions The maintenance work activities were performed well. The completed surveillance tests were verified to have met the test acceptance criteria. The inspectors concluded that these activities were performed in an acceptable manner.
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M3 Maintenance Procedures and Documentation H3.1 Procedure Discrepancy a.
Inspection Scope (62707)
The inspectors reviewed a letter dated July 19, 1996, from the plant manager to the NRC. The letter described the licensee's actions to modify the reactor building superstructure. The inspectors also reviewed applicable TS and UFSAR sections and WO 9601980, " Install Reinforcing Steel Reactor Building Roof and Columns."
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Observations and Findinas
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The inspectors noted that page 4 of Attachment A of the letter stated that the spent fuel pool (SFP) radiation monitor would be bypassed and/or removed when construction activities were in the vicinity of the detector.
The inspectors confirmed that this action was not a violation of TS or the UFSAR.
However, the inspectors identified that step 8.8.1
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of WO 9601980 allowed removal of the channel "A" SFP radiation monitor
and bypass the reactor building plenum (RBP) radiation monitor. The
licensee also planned to bypass the channel "B" radiation monitors later
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in the process.
Each channel would be inoperable for about 3 days.
Although bypassing the RBP radiation monitor was allowed by TS, the
inspectors determined that the action was not necessary.
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The inspectors discussed this observation with the design supervisor and
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system engineer. The system engineer stated that the intent of the bypass was not clear in the work order. The system engineer originally requested that the RBP radiation monitor be bypassed during the removal of the SFP radiation monitor. This would prevent a spurious Group 2 containment isolation. The RBP radiation monitor was to be restored after the spent fuel pool monitor was re-installed. The system engineer revised WO 9601980 to reduce the RBP radiation monitor inonerability
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time. A condition report was initiated to determine why the system engineer's bypass sequence was not translated clearly into the work
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order.
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Conclusions The inspectors determined that the original work order was adequate.
The system engineer promptly revised the work order to reduce the plenum monitor out-of-service time.
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M7 Quality Assurance in Maintenance Activities
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M7.1 Use and Documentation of Overtime Deviations
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Inspection Scone (71707)
The inspectors reviewed the following documents:
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j security ingress and egress times for the months of April and May
1996 for 15 individuals. These records aided in determining
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potential deviations from overtime guidelines. The inspectors
recognized that this information did not necessarily reflect actual work hours.
i Forms 3361, " Authorization to exceed Overtime Work Restrictions,"
l completed for January to June 1996; 4 Administrative Work Instruction (AWI) 08.10.01, " Overtime
j Restrictions and Fitness for Duty Requirements;"
Quality services department finding, FG-94-50, " Technical
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Specification Overtime Requirements;"
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Technical Specification 6.1.F; and
l Quality services department finding, FG-96-20, " Procedure
Adherence--0vertime Work Restrictions."
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Observations and Findinas The inspectors noted that AWI 08.10.01 was more conservative than TS 6.1.F by placing overtime restrictions on most individuals involved with safety or non-safety related work. Technical specification required limitation on personnel performing safety-related work.
The inspectors observed that during non-outage periods few deviations from the overtime guidelines occurred.
The inspectors noted that deviations from the overtime guidelines were not excessive during the refueling outage.
In most of these incidents, supervisor approvals were obtained prior to an individual exceeding guidelines. However, the inspectors noted that construction personnel routinely received approval days after a deviation occurred.
Deviation forms were completed for the group; not on an individual basis as required by the procedure. The inspectors reviewed these lists and determined that most of the individuals worked about 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> within a 48-hour period. This work included erecting and taking down scaffolding and other support work for safety and non-safety-related work orders.
The inspectors discussed this issue with the plant manager who stated that the practice did not meet management's expectations. The inspectors noted that a quality services auditor identified the same concern in observation report 1996238.
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The inspecters also observed the following:
Supervisors were only required to identify which guideline was
exceeded. The total hours worked by the individual was not recorded.
It was difficult to ascertain the degree of deviation and the appropriateness of corrective actions.
Some system engineers were onsite for excessive periods. One
individual was within the protected area for more than 93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br /> within a 7-day period.
The superintendent of engineering stated that tests and maintenance activities were often delayed a few hours. Some individuals preferred to rest onsite while waiting for the activities. The inspectors did not observe fatigue-related concerns during the outage for the individual in question.
j Some poor documentation examplos were observed. These included
not documenting limitations on work or plans to comply with overtime restrictions. The inspectors noted that the quality services auditor also identified this concern.
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Conclusions The licensee adequately implemented controls on overtime. The inspectors did not identify a violation of TS 6.1.F.
The administrative discrepancies identified were also identified by the licensee's auditing group and were being addressed through the corrective actions program.
III. Encineerina El Conduct of Engineering El.1 Observations of Operations Committee Meetinas a.
Insoection Scope (40500)
The inspectors observed several operations committee (0C) meetings.
Documents reviewed to prepare for the meetings included licensee draft copies of:
Safety Review Item (SRI)96-031, "HELB Assessment of "A" RHR Room
with Barrier HELB-2 Removed;"
SRI 96-0033, " Evaluation of RWCU Piping High Energy Line Break,"
revisions 0 and 1; (
Procedure 0122, "96 Hour Reactor Coolant I-131 Dose Equivalent
l Activity;" and l
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Design Change 96Q015 (Part C), " Control Room Dose Reduction -
Install Permanent Blanking Plates for the Control Room Intake
Duct."
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Qhservations and Findinas j
i The inspectors observed that the OC members were prepared and asked probing technical questions to the material presenters. The inspectors
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also made the following observations:
The licensee initiated SRI 96-031 to support maintenance on the
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- 11 RHR pump. A HELB barrier needed to be removed to facilitate lifting of the pump motor. Engineering personnel presented
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preliminary conclusions based on non-approved calcolations of the HELB affects. The inspectors observed that the OC members
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i acknowledged that SRI 96-031 could not be approved until the final calculations were received and reviewed.
The committee discussed the implications of not approving the SRI and decided to postpone
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l the RHR work until the final calculations were approved.
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A step was added to procedure 0122 to direct operators to inform
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j management if the I-131 concentration was greater than 0.25
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microcurie per gram. The OC members approved the procedure,
however, the inspectors noted that the newly added step would not
j be invoked by the operators since the previous step directed the l
operator elsewhere in the procedure. The inspectors notified the chemistry supervisor of this oversight. The error was promptly i
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The OC members identified that design change 96Q015 assumed a fan
in the control room lavatory was operating continuously. The OC
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members acknowledged that positive controls to ensure continued
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i operation were necessary. The inspectors verified that the such i
controls were implemented by the system engineer.
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Conclusions 4~
j The inspectors noted that the OC members thoroughly reviewed the subject
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material. The decisions made by the committee were technically sound and demonstrated conservative operation.
E2 Engineering Support of Facilities and Equipment E2.1 Discrepancy with Hiah Enerav Line Break Analysis for Reactor Water Cleanuo System a.
Inspection Scone During reviews associated with the Monticello power rerate project, the licensee contracted General Electric to reperform a RWCU HELB analysis.
On August 13, 1996, the licensee notified the inspectors that the
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analysis discussed in the UFSAR Section 5.3.5 was incorrect. This section of the UFSAR discussed an analysis of the secondary containment
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However, the licensee identified that at power levels less than 77
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percent, a break in the RWCU system would be more limiting.
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The licensee determined that the radiological consequences of the RWCU HELB were also greater than previously assumed in UFSAR Chapter 14.
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Additionally, the thermal-hydraulic conditions for some areas of the l
reactor building worsened in that some areas previously considered to be
mild environments became harsh environments.
The inspectors evaluated the licensee's corrective actions associated
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with the incorrect accident analysis.
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Observations and Findinos l
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General Electric personnel reperformed the RWCU HELB analysis assuming
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differing initial reactor power conditions. The re-analysis showed that i
at 77 percent power, the mass release was 443,260 lbm. The original
mass release of 168,400 lbn was incorrect.
The licensee determined that at power levels above 87 percent, the mass release rate was bounded by that assumed in the main steam line break analysis. Additionally, the re-analysis showed that at these power levels, an assumed RWCU break would automatically isolate due to low reactor water level. At lower power levels, an automatic isolation could not be assumed since reactor water level would not decrease to the low level setpoint.
Excess reactor feed flow would make up the lost inventory through the break and maintain water level above the low reactor water level.
Since automatic isolation was not ensured, the licensee performed an analysis and assumed operators would manually isolate the system after 10 minutes. This analysis showed that the most limiting condition would exist at 77 percent power.
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In response to this issue, the licensee initiated the following actions:
Operators isolated the RWCU until further reviews were complete.
- Engineering personnel performed an evaluation to justify continued
operation. The evaluation allowed operation to continue without i
restrictions at power levels above 90 percent. Operators were i
required to isolate the RWCU system at power levels below 90 percent except during startups and shutdowns. The licensee's technical basis relied on (1) manual operator action to recognize a RWCU line break by monitoring the RWCU room temperature and area radiation alarms; (2) existing alarm response procedures; and (3)
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operator training. The licensee justified not isolating RWCU
during startups and shutdowns since plant operation in these modes i
was less than 2 percent of total operating time.
This evaluation
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was approved by the OC on August 16.
The inspectors reviewed the evaluation and were concerned that no time restrictions were placed on startups and shutdowns. The inspectors
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discussed this concern with the plant manager. The evaluation was
revised to require plant manager's approval before unisolating the
system at power levels less than 90 percent.
The operators restored i
RWCU to service when the evaluation was complete.
The inspectors had no i
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further concerns.
l On August 22, engineering personnel completed another evaluation that eliminated all restrictions on RWCU system operation.
This evaluation
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indicated that with initial reactor water coolant dose equivalent i
iodine-131 concentrations less than 0.25 microcurie per gram, the
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radiological consequences for a RWCU HELB were bounded by the original
UFSAR analysis for a main steam line breek. This iodine-131
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i concentration was significantly less than the original assumed value; J
however, it was greater than normal operating values. The licensee
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revised chemistry procedures to reflect this new concentration limit.
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i The inspectors' initial review determined this evaluation was
acceptable. However, this is considered an Unresolved Item (50-
263/96007-03) pending NRR review of the justification to change the L
assumed iodine concentrations and any associated TS amendments.
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Conclusions
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The inspectors concluded that the licensee's August 16 evaluation was
acceptable. The licensee's August 22 evaluation is still under review.
The licensee identified additional actions such as UFSAR revision and j
potential modifications to correct the problem may be required.
Completion of these actions are considered an Inspection Followup Item (50-263/96007-04).
E2.2 Emeroency Service Water Pumo in the ALERT Ranae j
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Inspection Scope (61726)
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The inspectors reviewed the August 4, 1996, results of surveillance test 0187-2, "#12 Emergency Diesel Generator (EDG)/ #12 Emergency Service Water (ESW) System Tests." The operators measured several pump performance parameters and determined the acceptance criteria for differential pressure was not met.
The inspectors also reviewed applicable sections of TS and UFSAR.
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i b.
Observations and Findinas
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The inspectors noted that operations personnel notified the system
engineer to evaluate pump operability. The system engineer prepared i
form 3108, " Pump, Valve, Instrument Record of Corrective Action," and
determined that the pump remained operable. This determination was
based on a review of previous tests. The engineer also recommended 1-testing within 6 weeks.
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The inspectors independently verified the surveillance results and i
calculations. The inspectors confirmed that the #12 EDG-ESW pump j
performed in the ALERT range for low differential pump pressure. The inspectors also reviewed previous test data including vibration information and did not observe a decline in the pump's performance.
The inspectors reviewed the engineering evaluation and had no concerns.
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Conclusions The inspectors determined that the operability evaluation and accelerated testing were appropriate actions. The system engineer provided a sound operability evaluation.
E2.3 Results of UFSAR Review A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures, and parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices, procedures, and parameters.
As discussed in paragraph E2.1, the licensee identified an error in UFSAR Chapters 5 and 14.
IV. Plant Support
$6 Security Organization and Administration S6.1 Routine Meetina with Security Manaaement On August 19, 1996, the inspectors met with the corporate security supervisor, site superintendent of security, and others to discuss the licensee's quarterly self-assessment report. The inspectors noted that corrective actions were initiated to correct equipment deficiencies in a
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timely manner. The inspectors also noted that the licensee responded
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promptly and aggressively to an industry concern regarding
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l confidentiality of operator licensing examinations. The licensee installed an independent security system within the training building as
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an extra measure to secure the examinations.
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F4 Fire Protection Staff Knowledge and Performance
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F4.1 Performance Durina Simulated Fire
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a.
Inspection Scone (71750)
The inspectors observed fire brigade response to a simulated on-site fire on July 29, 1996. The inspectors also attended the post-drill
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critique.
b.
Observations and Findinas i
The simulated " fire" was caused by an overturned oil truck near the
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service building. This resulted in oil and fire entering the service
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building basement. The normal fire brigade assembly point and gear were not accessible.
l The inspactors observed good initial response by control room personnel.
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The operator immediately notified the shift manager of the " fire" and i
alerted fire brigade members. The shift supervisor assumed the role of
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i the fire brigade leader (FBL) and quickly evaluated the situation. The FBL notified security personnel to evacuate a post due to the proximity of the " fire." The security officers remained at this post, though, during the drill.
The inspectors noted that control room personnel promptly notified the fire department and other government agencies. Comunication between the FBL and Monticello Fire Department personnel was considered good.
Security personnel maintained sight observation of unbadged individuals.
The inspectors noted good discussion during the critique. The effects of the fire (i.e., heat, water drainage, etc.) were difficult to simulate during the drill; however, actions to address these issues were discussed. The drill leaders asked probing questions to ensure that the potential hazards and subsequent actions were understood.
Fire brigade members and Monticello Fire Department personnel recomended actions to improve comunications and response times.
c.
Conclusions The fire brigade and control room personnel responded promptly and appropriately during the fire drill. The inspectors noted good comunication between the site and Monticello Fire Department personnel.
The licensee's critique was thorough.
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F5 Fire Protection Staff Training and Qualification
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F5.1 Lic3niee Identified Discrepancy with Resoiratory Mask Fits
a.
Insoection Scope (71750)
On August 7, 1996, the general superintendent of radiation services notified the inspectors that three of six radiological protection technicians fire brigade respirator fit qualifications had lapsed in June 1996. The licensee initiated condition report 96001857,
"Inoividuals Respirator Qualifications Have Lapsed." The radiation protection supervisor notified operations and chemistry personnel of the j
discrepancy.
b.
Observations and Findinas The inspectors discussed the event with operations personnel and confirmed that operations and chemistry personnel with fire brigade
responsibilities were properly qualified. The inspectors confirmed that the individuals received respiratory mask fits before resuming fire brigade duties.
Technical specification 6.1.C.6 required that five individuals qualified as fire brigade members be onsite at all times. Preliminary i
investigation showed that additional personnel qualified in fire brigade duties were available when these individuals were onsite. This is
considered an Unresolved Item (50-263/96007-05) pending inspectors'
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confirmation of TS compliance and review of the licensee's corrective actions.
c.
Conclusions The radiation protection supervisor responded promptly to ensure all
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individuals assigned fire brigade responsibilitier e s properly qualified.
V. Manaaement Meetinas X1 Exit Meeting Sumary
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On August 28, 1996, the inspectors presented the inspection results to me..a es of licensee management. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the
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inspection should be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED Licensee E. Watzl, Vice President Nuclear W. Hill, Plant Manager M. Hammer, General Superintendent Maintenance K. Jepson, Superintendent, Chemistry & Environmental Protection L. Nolan, General Superintendent Safety Assessment M. Onnen, General Superintendent Operations E. Reilly, Superintendent Plant Scheduling C. Schibonski, General Superintendent Engineering W. Shamla, Manager Quality Services J. Windschill, General Superintendent, Radiation Protection L. Wilkerson, Superintendent Security B. Day, Training Manager INSPECTI0h PROCEDURES USED IP 37550:
Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations IP 71707:
Plant Operations IP 71750:
Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 50-263/96007-01 NCV Failure to perform correct revision of surveillance 50-263/96007-02 NCV Failure to have procedures for recirculation pump 50-263/96007 03 URI NRR review of revising iodine concentration and TS change 50-263/96007-04 IFI Completion of UFSAR changes and modification to resolve issue.
50-263/93007-05 URI Followup on licensee actions with regard to respiratory mask fit and fire brigade composition Closed 50-263/96006-01 URI Licensee-identified procedure was not reviewed properly
LIST OF ACRONYNS USED AWI Administrative Work Instruction CFR Code of Federal Regulations CS Core Spray ECCS Emergency Core Cooling System EDG Emergency Diesel Generators ESW Emergency Service Water FBL Fire Brigade Leader HELB High energy line break IFI Inspection Followup Item MG Motor Generator NCV Non-Cited Violation NRC Nuclear Regulatory Commission OC Operations Committee RBP Reactor Building Plenum RHR Residual Heat Removal System RWCU Reactor Water Clean-Up SFP Spent Fuel Pool TS Technical Specification UFSAR Updated Final Safety Analysis Report URI Unresolved Item W0 Work Order i
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