IR 05000263/1998016
| ML20198B790 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/14/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20198B779 | List: |
| References | |
| 50-263-98-16, NUDOCS 9812210125 | |
| Download: ML20198B790 (19) | |
Text
. - .. .- . . _._.- .- - . - _ . - -
. <
U. S. NUCLEAR REGULATORY COMMISSION
.
REGIONlli Docket No: 50-263 License No: DP.R-22
,
Report No: 50-263/98016(DRP)
.
Licensee: Northern States Power Company l Facility: Monticello Nuclear Generating Station Location: 2807 West Highway 75 i
,
Monticello, MN 55362 1 Dates: October 14 through November 30,1998 i Inspectors: S. Ray, Acting Senior Resident inspector
D. Wrona, Resident inspector 1
.
Approved by: B. Burgess, Chief Reactor Projects Bran ?
!
l
.
9812210125 981214 PDR 0 ADOCK 05000263 PM
.
I
!
EXECUTIVE SUMMARY Monticello Nuclear Generating Station l Inspection Report 50-263/98016(DRP) !
This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspectio !
Operations
.
In general, the plant was operated in a controlled and deliberate manner. Operators were knowledgeable of current and planned activities and acted in a professional l manner. The inspectors noted one minor instance of an operator not highlighting a '
discrepant log entry as recommended by station procedures. (Section 01.1)
.
The rerate power ascension testing program was conducted in a careful, deliberate manner. The associated test procedure was well written and properly followe Numerous licensee groups provided timely support of the program. (Section 01.2)
.
A power reduction to allow maintenance on a recirculation pump motor generator set was performed well with a thorough pre-evolution briefing, close supervision, and smooth reactivity manipulations. (Section 01.3) {
'
.
Material condition of the 125-volt direct current electrical system and the high pressure coolant injection system was good. The system engineers were cognizant of the current condition and concerns associated with their respective systems. No concerns were identified by the inspectors. (Section O2.1)
l Maintenance
. In general, the observed maintenance and surveillance activities, which involved systems or components, such as the high pressure coolant injection system and the 125-volt direct current electrical system, were conducted in accordance with procedures and in a professional manner. (Section M1.1)
i Enaineerina i
'
. Engineering personnel support to operations and maintenance activities, including the attempt to identify the source of increased leakage in the drywell, was appropriate during this inspection period. (Section E1.1)
Plant Support
. During performance of an emergency preparedness drill, notifications were promptly made and technical support center staffing was timely. Also, good team work was noted between the various emergency response groups. A minor instance of a sen:or member of an organization not relieving a junior member as recommended by station procedures was noted. (Section P1)
. . .~ .- . . . _ . _ .~ -~_. . _ - - - - = _ _ - . . = . .-
. .
b
.. ' The fire brigade training requirements of plant procedure 4 AWi-08.01.01 were not correctly implemented. A review by plant personnel of a similar finding at Prairie Island ,
could have identified this issue earlier. A Technical Specification violation for a failure to l follow a fire protection procedure was cited.- (Section F5)
i
!
l l ,
i
!
!
!
!
i
.l .
. . .
. _ . . . . . . _ _ _ . _ _ _ . __ __ _ _ _ _ - . _ _ _ . _ _ . ___ . _
. .
t- ;
Report Details
[ J l
,
Summary of Plant Status l
From the ~ o eginning of the inspection period until October 21, the unit operated between 85 and 100 percent power while testing associated with a power rerate was conducted. On '
, October 21, power rerate testing was completed and the unit was returned to 100 percent power. On October 22, the licensee initiated power accession to the new licensed thermal power, and the new 100 percent power was attained on October 30. On November 3, the l
operators reduced power to 50 percent to make repairs to the 11 recirculation pump motor generator set. The unit operated at 100 percent for the remainder of the inspection period.
i
-
1. Operations
. 01 Conduct of Operations
!
l 01.1 General Comments Insoection Scope (Inspection Procedure (IP) 11_70J7J i l
!
The inspectors conducted frequent reviews v ongoing plant operations, including
'
observations of control room evolutions, shift turnovers, and operator rounds. The inspectors also reviewed control room logbooks and operability determinations.
Updated Safety Analysis Report (USAR) Section 13, " Plant Operations," was reviewed
- as part of the inspection.
. Observations and Findinas in general, the plant was operated in a controlled and deliberate manner. Specific events and noteworthy observations are detailed in the sections below. Operator ;
"
performance during roviine operations and surveillance test activities was goo , Operator turnovers were conducted in accordance with procedures. Operators were 1 knowledgeable of current and planned activities and acted in a professional manne Since a hat rack was installed outside of the control room, the inspectors noted that operators have not walked by emergency core cooling system control panels while
'
wearing hardhats, as they had previously, as discussed in inspection Report 50-263/98012(DRP). The inspectors also noted one minor instance of an operator not highlighting a discrepant log entry as recommended by station procedures.
l Conclusions
,
in general, the plant was operated in a controlled and deliberate manner. Operators
-
were knowledgeable of current and planned activities and acted in a professional manner. The inspectors noted one minor instance of an operator not highlighting a discrepant log entry as recommended by station procedure s
,
.- ._ . - - - _ _ - --. . _ -
_ -
. . .
.
01.2 Rerate Power Ascension Testino Prooram Inspection Scope (IP 71707)
The inspectors observed major portions of the rerate power ascension testing program performed in accordance with Special Procedure 8303, " Prerequisites and Power Ascension Testing Control Procedure," Revision 0. Observations included pre-evolution briefings, transient testing, power changes, and Operations Committee (onsite review committee) meeting Observations and Findinos As discussed in Inspection Report 50-263/98015(DRP), Section 01.4, the licensee received a license amendment to allow an increase in licensed maximum power from 1670 to 1775 megawatts-thermal (MWt). The licensee started its power ascension testing program on October 8,1998. Between the beginning of this inspection period on October 14,1998, and October 21,1998, the plant was operated at various power levels between 1503 and 1670 MWt. This was between 90 and 100 percent of the previously licensed maximum power. Various transient testing was completed to benchmark plant performance and gather data to predict parameters at higher powers. Transients included small step changes to reactor water level, steam pressure, and recirculation flow control setpoint On October 21,1998, testing at up to 1670 MWt was completed and the results were reviewed and accepted. The Operations Committee released the plant for operations above 1670 MWt. Using Special Procedure 8303, the licensee then raised reactor power to 1775 MWt in three equal steps on October 22,26, and 30. Prior to each step, engineers predicted what various plant equipment parameters would be at the next level and held a detailed briefing for the control room and outplant operators. At the completion of each step, equipment was carefully monitored until new steady-state operating characteristics were determine At 1740 MWt, plant staff raised concerns about the performance of the feedwater regulating valves. In addition, the capacity of the feedwater pumps to provide adequate flow margin to be able to respond to level change demands at 1775 MWt was questioned. The engineers developed an attachment to Special Procedure 8303 directing additional dynamic testing consisting of a gradually increasing series of water level transients to verify the proper performance and capacity of the feedwater syste Those tests were successfully completed on October 29,199 The plant reached the final maximum licensed power level of 1775 MWt for the first time on October 30,1998. The higher power resulted in an increased electrical output of about 40 megawatts as expected. All plant equipment performed at the higher output levels within design parameter All of the power ascension testing steps were performed carefully, deliberately, and in accordance with procedures. Informative briefings were held prior to each major ste In the briefings, expected equipment performance, duties of each individual, and
_
_ _ _ . _ _ - _ _ _ _ _ _ _ . ._ ...__.._ _ _ __ _ .
contingency plans for unexpected problems were discussed. Equipment logs, transient co:nputer point traces, and other appropriate data were reviewed in detail at each step.
i Special Procedure 8303 was well written and easy to understand. Temporary changes L te Special Procedure 8303 were made as needed in accordance with Technical Spec!Retion Licensee quality services inspectors observed performance of all of the significant steps of the test program. Support by system engineering, nuclear engineering, health physics, instrument and controls, and many other groups was timely.
, Conclusions The rerate power ascension testing program was conducted in a careful, deliberate manner. The tast procedure was well written and properly followed. Numerous licensee i groups provided timely support of the program.
l 01.3 Power Reduction for Work on Recirculation Pumo Motor Generator Set jarection Scooe On November 3,1998, the operators reduced reactor power to allow replacement of a brush on the 11 recirculation pump motor generator set. The inspectors observed the power reduction and subsequent ascension back to full power. Operations Manual Section C.2, " Power Operation," Revision 10, was reviewed as part of this inspection, Observations and Findinas The purpose of the power reduction was to place the plant in a condition that would prevent entry into the instability region of the operating curves if the recirculation pump tripped during the maintenance activity. Before the power reduction, a briefing was held for the maintenance personnel, operators, and system engineer involved. The briefing covered the purpose for the evolution, precautions, expected plant response, personnel assignments, and communications equipment. In addition, a 1985 event in which the recirculation pump tripped during a similar job was discussed. The briefing was informative and thoroug Performance of the power reduction was controlled well. The rate of power change had to be closely monitored and smoothly controlled to prevent exceeding temperature j- rate-of-change limits for the circulating water discharge canal. Operators closely l followed a memorandum from the nuclear engineers regarding the coordination of l recirculation flow and control rod movements. They were ele to make the power reduction expeditiously without exceeding the discharge lim:ts. The power reduction was almost complete when the next operating crew reported for work, but the on-duty i;' reactor operators completed all of the control rod manipulations before relinquishing control of the reactor to the oncoming cre , -
g - y -- -'" * - -
.. . . . . - - . - . - - . - . - - - - - _ _ _ . - - - - - --
. .
I The maintenance work took less than one hour and the reactor was retumed to full :
power shortly thereafter. The entire evolution was performed very smoothly, with close supervision by the shift supervisor and shift manager, and in accordance with the applicable procedur Conclusions '
- A power reduction to allow maintenance on a recirculation pump motor generator set L was performed well with a thorough pre-evolution briefing, close supervision, and i L
'
smooth reactivity manipulation )
02 Operational Status of Facilities and Equipment
'
O2.1 Enaineered Safety Feature Systen? Walkdowns Insoection Scope (IP 71707)
The inspectors used recent events and the licensee's risk analysis to aid in the determination of which safety related systems to walk down and selected the following systems:
I*
High pressure coolant injection (HPCI) system
- 125 volts direct current (Vdc) system Observations and Findinas
.
A walkdown of the HPCI system was conducted following the quarterly run of the J HPCI system due to the minor water hammer experienced as discussed in l Section M1.1. No concems were note !
-. . The inspectors walked down the 125-Vdc system due to its high risk i achievement worth importance measure ranking as identified in the licensee's ;
'
risk analysis. The following documents were reviewed:
- Technical Specification (TS) Sections 3.9.B.4 and 4.9.B.4;
= USAR Section 8.5;
,
'
- Operations Manual Section B.9.10, "125 Vdc System";
. Battery Technical Manual NX-16647, "C&D Batteries," Revision 4;
- Drawing NX-41583-1, "Monticello Nuclear Generating Plant 11 Battery Arrangement," Revision A;
= Drawing NX-41583-2, "Monticello Nuclear Generating Plant 12 Battery Arrangement," Revision A; and
- Drawing NX-41583-3, "Monticello Nuclear Generating Plant Battery Arrangement 2-Tier & 2-Step Racks," Revision :
.
!
The 125-Vdc system batteries were installed in accordance with drawings NX-41583-1,2 and 3. The system engineer was aware of the current status of the batteries and reviewed the weekly battery surveillance procedures. No concerns were note Conclusions Material condition of the observed systems was good. The system engineers were cognizant of the current condition and concerns associated with their respective systems. No concems were identifie Miscellaneous Operations issues (IP 92901)
O8.1 (Closed) Violation (VIO) 50-263/98002-01(DRP): This violation involved the failure to provide adequate instructions for the restoration of MO-2078. Specifically, the restoration instructions did not contain instructions to require the operators to manually unseat the valve and position the manual declutch after the valve was manually shu The licensee has revised Administrative Work Instruction (4 AWI) 04.04.01, " Equipment Isolation," Revision 14, and 4 awl-04.04.02, " Equipment Positioning, Witness Check, and independent Verification Methods," Revision 4, to add guidance for isolating motor-operated valve (MOV) hand wheels and to add requirements for manually positioning MOVs. Training was provided to operations personnel on lessons-leamed from this even .2 (Closed) VIO 50-263/98002-02(DRP): This violation involved the failure to remove the hold-and-secure card from the #13 residual heat removal service water (RHRSW) pump hand switch prior to declaring the pump operable. The inspectors verified the following procedures included revisions associated with the use of the Shift Supervisor's Hold and Secure Card Book:
.
Procedure 1047-1, " Shift Supervisor Checklist," Revision 8; a
4 awl-04.04.01, " Equipment isolation," Revision 14;
.
owl-01.06, " Duty Operations Personnel Requirements and Responsibilities,"
Revision 1; and
.
owl-03.02, " Safety Related System Operability," Revision This violation was also discussed in operations training which reiterated operability verification considerations and plant status awarenes _ _
. _ _ . _ _ . - _ _ _ __ _ _. _ _ __ _ _ _ .. __ _ _ ._ _ _ _ _ _ . . _
!
<
11. Maintenance i M1 Conduct of Maintenance
! M1.1 General Comments j Insoection Scooe (IPs 62707 and 61726)
'
The inspectors observed all or portions of selected maintenance and surveillance
!
activities. As practical, the inspectors selected maintenance and surveillance activities associated with systems that were risk significant. Included in the inspection was a !
review of the surveillance procedures and work orders (WOs) listed, as well as the i
,
appropriate USAR sections pertaining to activities inspected.
- Observations and Findinas
.
I In general, the inspectors observed that the work associated with these activities was
! conducted in a professional and thorough manner. All work observed was performed j with the work package present and in active use. Technicians were experienced and
- knowledgeable of their assigned tasks. The inspectors frequently observed supervisors
] and system engineers monitoring job progress, and quality control personnel were present whenever specified in procedure. When applicable, appropriate radiation control measures were in-plac The following maintenance and surveillance activities were observed:
.
Surveillance Procedure 0194, "11 and 12,125 Vdc Battery Operability Check,"
Revision 11,
. WO 9803221 "HPCI High Steam Flow Isolation,"
The inspectors noted that the instructions specified in WO 9803221 to replace DPIS23-768 were more detailed than those specified in WO 9802698,
"DPIS23-76B Bellows Fill Liquid Leak," which was discussed in Section M1.1 of Inspection Report 50-263/98012(DRP). The instructions in WO 9803221 specified wire termination points and independent verification .
Procedure 7130, "HPCI System Instrument Maintenance Procedure,"
Revision 12,
. Surveillance Procedure 0255-06-IA-1, "HPCI System Test with Reactor Pressure at Rated Conditions," Revision 40, The HPCI system engineer was stationed in the HPCI room during the cycling of MO-2068 and instrumentation was installed on HPCI piping to observe for a possible water hammer. During the cycling of MO-2068, a minor water hammer disturbance was noted, Tnis disturbance was smaller than some of the previous water hammers experienced. The system engineer walked down the HPCI
system to verify operability. The inspectors independently walked down the HPCI system and no concerns were identified. Water hammer issues associated with the HPCI system are also discussed in Section E *
Procedure 0248, " Reactor Building Vent Wide Range Gas Monitor Calibration,"
Revision 18,
.
WO 98903342, " Tighten Packing to Stop Leakage [for various control rod drive valves),"
.
Surveillance Test 0003, "Drywell High Pressure Scram and Group 2, 3, and SCTMT [ secondary containment] Isolation Test and Calibration Procedure,"
Revision 14,
.
Surveillance Test 0081, " Control Rod Drive Scram Insertion Time Test,"
Revision 29, and
.
Special Procedure 8303, Attachment H, " Reactor Water Level Dynamic Testing,"
Revision Conclusions
.
In general, the observed maintenance and surveillance activities, which involved systems or components such as HPCI and batteries, were conducted in accordance with procedures and in a professional manner. Engineering staff provided good support to the maintenance staf M8 Miscellaneous Maintenance issues (IPs 92902 and 92700)
M8.1 (Closed) Licensee Event Report (LER) 50-263/98-002: Main Steam isolation Valve Position Setpoint Outside Allowed Rang This event involved the licensee's discovery, during a refueling outage, that seven of the eight valve position switches for the inboard main steam isolation valves (MSIVs) had setpoints outside of the Technical Specifications required band. The discovery was believed to be due to the use of a new, more accurate, surveillance technique along with switches with too large of a deadband compared to the length of the valve stroke. The licensee replaced all of the switches with an improved mode The event was considered to be of low safety significance because the redundant limit switches on the outboard MSIVs all were found to be within tolerance. In addition, although the setpoints of the seven switches were out of the TS band, they were all found to be within the value assumed in the accident analysi __. . ._ __ _ _ _ . . _ . .. .
Although the licensee discovered the condition during a refueling outage, it assumed that the condition had existed during some of the previous operating period, as reported in the LER. However, because the inspectors found no indication that the plant had !
actually operated while the setpoints were outside of the Technical Specification limits, i no violations were identifie l l
M8.2 (Closed) LER 50-263/98-005: HPCI Removed from Service to Repair Steam Leak in !
Drain Trap Bypas !
This event was previously discussed in Inspection Report 50-263/98015(DRP), l Sections 01.1.b and M2.2. The leaking air-operated valve was replaced with a manual '
valve under the licensee's temporary modification program. The licensee intended to replace the manual valve with an improved air operated valve at a future time. The j inspectors had no concerns with this even Ill. Enaineerino i l
E1 Conduct of Engineering I
,
E General Enaineerina Observations Inspection Scope (IP 37551)
l The inspectors reviewed engineering-related activities and observed engineering personnel involvement in resolving problems identified during maintenance, surveillance, and operations activities. Included in the inspection was a review of appropriate TSs and USAR section Observations and Findinas F! ant p ~Annel identified an increasing trend in the unidentified drywell leakrat Although the leak rate was approximately 0.31 gallons per minute (gpm), which is well below the TS limit of 5 gpm, the licensee formed a team composed of engineering, scheduling, chemistry, radiation protection and operations personnel to attempt to identify the source of the leak. Condition Report 98003050 was written to track this issu Conclusions Engineering personnel support to operations and maintenance activities, including the attempt to identify the source of increased leakage in the drywell, was appropriate ;
during this inspection perio !
\
I l
I l
!
.. - . . - - - - . .. . --_ __ - - - -. . . -
i
\
l E8 Miscellaneous Engineering issues (IPs 92700 and 92903)
- E8.1 (Closed) Inspection Follow-uo item (IFI) 50-263/98007-01(DRP): Water Hammer Event in HPCI Syste This issue was previously discussed in Inspection Reports 50-263/98004(DRP),
Section E1.1, and 50-263/98007(DRP), Section E2.2, and involved the occurrence of
,
minor water hammers when the HPCI system was started for surveillance testing. The
'
inspectors reviewed the status of the licensee's investigation and corrective actions to date, as recorded in Condition Report 98000413, and interviewed the system enginee In addition, the inspectors noted that the licensee's Safety Audit Committee discussed the issue in their most recent meeting.
'
Although the licensee had not conclusively determined the root cause of the water hammers, it had determined that the most likely cause was the buildup of a steam void ;
.
in the piping due to slight seat leakage on injection isolation valve MO-2068. Monitoring :
!
of the void size, using ultrasonic detection equipment and piping temperature measurements, indicated that the size of the void was stable and the situation was not
degrading. Inspections of the piping system did not indicate any significant pipe ,
movement or damag I
, The licensee contracted a consulting firm to investigate the issue and make
'
recommendations for long-term corrective actions. A purchase order for a new MO-2068 valve was issued. The inspectors had no operability concerns with the system and determined that the licensee's investigation and corrective action process would be
,
adequate to resolve the problems.
-
E8.2 (Closed) IFl 50-263/98011-01(DRP): Elevated Steam Chase Temperatures.
i This issue was previously discussed in Inspection Report 50-263/98011(DRP),
Section 01.2, and involved concerns with elevated temperatures in the main steam chase and the effect on environmental qualification of equipment. The insoectors
- reviewed Condition Report 98001727 and discussed the issue with the environmental
qualification system engineer. The licensee believed the elevated temperatures were probably related to leakage across HPCI valve MO-2068 as discussed in Section E of this report. The situation appeared to have stabilized and steam chase temperatures have exceeded 135 degrees (Fahrenheit) only occasionall i The licensee determined that the 135 degree environmental qualification temperature was not a limit, but was the average temperature for which equipment in the steam chase was qualified. Since the average temperature over the lifetime of the equipment was well under 135 degrees, the inspectors had no additional operability concerns. The steam break accident analysis assumed that initial steam chase temperature was i 140 degrees.
The licensee performed analysis that demonstrated that the peak pressure and temperature in the steam chase during a steam break accident was relatively insensitive
12
1
_ _ ._ _ __ _ __ _ _ . . _ _ _ _ . _ . . . _ _ - _ _ - _ - _ _
. .
l to initial room temperature. The licensee performed an analysis that assumed an initial temperature of 165 degrees and showed that design basis limits would still be met.
,
!
As an interim compensatory measure, the licensee issued temporary change 1708 to Abnormal Procedure C.4-B.8.7.A, " Ventilation System Failure," Revision 8, which l
contained instructions for hourly logging of steam chase temperatures if they exceeded L 135 degrees and reduction of reactor power if temperatures could not be maintained ;
j below 165 degree l As discussed in Inspection Report 50-263/98011(DRP), one of the inspectors' concerns I was that operators did not initially initiate any compensatory measures wher. they {
T recorded steam chase temperatures above the acceptance limit on their log sheet l The inspectors have noted a significantly increased awareness on the part of the i operators to close monitoring of steam chase temperatures since that even j t
IV. Plant Suncort R1 Conduct of Radiological Protection and Chemistry Controls (IP 71750) l
! During normal resident inspection activities, routine observations were c.enducted in the
'
area of radiation protection. The inspectors noted that effective radiological controls were established and that technicians provided adequate support during maintenance I and sr.willance activities. No concerns were noted.
'
'P1 Conduct of Emergency Preparedness Activities Insoection Scooe (IP 71750)
On November 11,1998, the licensee performed an emergency preparedness exercis ,
The inspectors reviewed the exercise, observed controller training, and observed the exercise from the Technical Support Center (TSC) and Operations Support Center, b. - Observations and Findinas The planned exercise consisted of a simulated leak in the drywell that required classification as an alert. The exercise was conducted off-hours, at approximately 7:30 The initial notifications to the state and local authorities were made within the required time limits.. The NRC was promptly notified following the notification of the state and local authorities. The inspectors noted that the TSC was promptly staffed. In general, the emergency response organization board, located near the entrance to the TSC, was used effectively. One instance of a more senior member of an organization not relieving i his junior counterpart was noted by the inspectors. This was inconsistent with the expectations (but not the requirements) of A.2-100, " Emergency Organization,"
Revision 43. l _ _ _ ,
. _ _ . . _ . . . _ _ _ _ _ - _ _ _ _ _ _ - _ _ . _ _ - - - - _ - - -
,
,
3-
.
'
The emergency director (ED) quickly established command and control. Noise levels remained relatively low. Written logs were maintained by the various groups within the i
TSC. The ED gave adequate waming of upcoming briefings, which were held periodically and were informative. Communications were clear and effective. Good
teamwork was noted, particularly between the operations, engineering, and
'
maintenance groups. A responsibility of the Operational Support Group, as delineated I in A.2-100, is to backup the control room on the Emergency Operating Procedures
} Through discussion with the operations support group leader and review of written notes i
and logs, the inspectors determined that the operations group was aware of status of
{ the EOPs. Since most plant equipment remained available to maintain reactor vessel J
_
level and pressure, and since the rate of change of parameters such as drywell pressure and temperature were slow, the operations support group leader was able to maintain an awareness on the status of the EOP !
. The inspectors also attended the post-exercise critique. Participants and controllers provided constructive comments. The inspectors concluded that while the exercise
- scenario was not technically challenging, in that plant radiation levels remained normal, i j no offsite dose assessments and calculations were required, and plant parameters were l l not changing rapidly, the exercise provided a test of communications between the l various groups and a test of off-hours staffing of the emergency response organization.
i
- Conclusions
'
Notifications were promptly made. TSC staffing was timely. Good team work was noted between the various emergency response groups. A minor instance of a more senior j member of an organization not relieving his junior counterpart as recommended by
-
station procedures was noted.
i S1 Conduct of Security and Safeguards Activities (IP 71750)
- During normal resident inspection activities, routine observations were conducted in the
- area of security and safeguards activities. No concems were noted.
'
F2 Status of Fire Protection Facilities and Equipment (IP 71750)
, During normal resident inspection activities, routine observations were conducted in the
! area of fire protection. No concerns were noted, except for the issue discussed below in l Section F5.
!
-
F5 Fire Protection Staff Training and Qualification l'
- Inspection Scooe (IP 71750)
,
.
The inspectors attended the licensee's fire protection audit exit meeting. At the
- meeting, the inspectors questioned whether the licensee gave fire drill participation i credit to control room operators and informed them that a similar issue was discussed at
Prairie Islan ;
.I
, ., , , - _
.- .- -.-,1- -- . -
. Findinas and Observations After the issue of giving credit to control room operators for fire brigade drills was raised by the inspectors, the licensee conducted a review of fire brigade drill records. The licensee informed the inspectors that if credit is not given to control room operators then not all of the fire brigade members had participated in the required number of drill Monticello Administrative Work Instruction 4 AWi-08.01.01, " Fire Prevention Practices,"
Revision 11, Paragraph 4.6.1.B stated: "Each brigade member should participate in each drill, AND SHALL participate in at least two drills per year."
A review of fire brigade drills indicated that several brigade members had not met the requirement to participate in the required number of fire brigade training drills. The licensee stated that thev gave credit for control room coverage during fire brigade drills for meeting the drill requirement. However, this does not meet the requirements for training as a fire brigade member. A review by plant personnel of a similar issue at Prairie Island (documented in Inspection Report 50-282/98012(DRS);
50-306/98012(DRS)) could have identified this issue earlie Technical Specification 6.5.A.6 required that detailed written procedures be prepared and followed for the implementation of the fire protection program. The failure to correctly implement 4 awl-08.01.01 is contrary to TS 6.5. (VIO 50-263/98016-01(DRP)).
The licensee determined the cause was a misinterpretation of the requirement to participate in a fire drill. Once the licensee determined that certain fire brigade members did not participate in the required number of fire drills, they made them ineligible for fire brigade duty. The licensee performed fire brigade drills that included the unqualified fire brigade members to ensure personnel were properly qualified prior to assuming fire brigade responsibilities. All fire brigade members who were required to participate in a fire drill, successfully completed the drill. The licensee wrote Condition Report 98002955, " Credit given to on-shift control room personnel for participation in fire brigade drills," nas planned to revise the current practice to ensure participation requires actual on-scene response to drills, and has planned to review other aspects of the associated regulation to ensure complianc Conclusion The fire brigade training requirements of 4 awl-08.01.01 were not correctly implemented. This issue could have been identified earlier by plant personnel had a similar issue at Prairie Island been reviewed. A violation was identifie i t
_ _ _ _ - _ _ _ _ _
. .- ~. . - ..- - ---. _ _ - . . . .
I
!
!
V. Manaaement Meetinas I- X1 Exit Meeting Summary On November 30,1998, the inspectors presented the inspection results to members of licensee management. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection ,
should be considered proprietary. No proprietary information was identifie )
i
.
- i i
i i
l I
- . !
l l
..
l l
l'
,
>
..
.. . . . - . . . .- - . - . - - - .-. - - - . . . .
. .
i
'
!
!
PARTIAL LIST OF PERSONS CONTACTED
, Licensee
!
. M. Hammer, Plant Manager ;
B. Day, General Superintendent Operations j K. Jepson, Superintendent, Chemistry & Environmental Protection j L. Nolan, General Superintendent Safety Assessment E. Reilly, General Superintendent Maintenance i S. Hammer, Acting General Superintendent Engineering ,
A. Ward, Manager Quality Services i L. Wilkerson, Superintendent Security J. Windschill, General Superintendent, Radiation Protection
,
l l l l
,
I
!
I I
!
!
l
17 s
. . _. _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ . . _.. _ _ _ _ . - ..
..
l
INSPECTION PROCEDURES USED t
'
IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: _ Plant Support .
IF 92700: Onsite Follow-up'of Written Reports of Nonroutine Events at Power Reactor Facilities .
IP 92901: . Followup - Plant Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering ll lTEMS OPENED, CLOSED, AND DISCUSSED
]
. Opened 50-263/98016-01(DRP) VIO Failure to meet fire brigade drill requirements of i
4AWi-08.01.01 Closed 50-263/98002-01(DRP) VIO Inadequate instructions for the restoration of MO-2078 50-263/98016-01(DRP) VIO Failure to meet fire brigade drill requirements of 4AWi-08.01.01 l
50-263/98002-02(DRP) VIO Failure to follow 4AWi-04.05.05 during the restoration of the #13 RHRSW pump 50-263/98-00 LER Main steam isolation valve position setpoint outside i i allowed range j 50-263/98-005 LER HPCI removed from service to repair steam leak in drain trap bypass 50 263/98007-01(DRP) IFl Continued investigation into the origin of the slight water l-
>
hammers in the HPCI injection piping 50-263/98011-01(DRP) IFl Operators did not anticipate increased steam chase area
,
temperatures and potential operation outside licensing ,
basis I i
Discussed j-None
!
'
t
'. 18 I
i
'
_ _ . - , , .-. . _ . _ ,,
V
-
i .
,y -
LIST OF ACRONYMS USED
'AWI Administrative Work Instruction CFR Code of Federal Regulations .
.DRP Division of Reactor Projects - ,
ED Emergency Director EOP Emergency Operating Procedure gpm gallons per minute
' HPCI High Pressure Coolant injection IFl: Inspection Followup Item IP Inspection Procedure LER- Licensee Event Report .
MOV Motor-Operated Valve .
MSIV Main Steam Isolation Valve
'
MWt . Megawatts-Thermal NRC - Nuclear Regulatory Commission
!-' NSP Northern States Power
- OWI Operations Work Instruction
.l
'
PDR Public Document Room RHRSW Residual Heat Removal Service Water SCTMT Secondary Containment -
'
TS Technical Specification
-
TSC Technical Support Center c USA Updated Safety Analysis Report
[ Vdc Volts-Direct Current VIO Violation WO Work Order s
'1 i
!
,
s
._
!
!
l l
l