IR 05000282/1997018
| ML20199G906 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/14/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20199G879 | List: |
| References | |
| 50-282-97-18, 50-306-97-18, NUDOCS 9711250332 | |
| Download: ML20199G906 (22) | |
Text
'
U.S; NUCLEAR REGULATORY COMMISSION
_
,
REGION 111 i
'
.
Docket No.
50 282, 50-306 l License No.
,
Report No.
50 282/97018(DRP); 50-306/97018(DRP).
Licensee:
Northern States Power Company
,
Facility:
.. Prairie Island Nuclear Generating Plant
- Location:
1717 Wakonade Drive East Welch, MN 55089 Dates:
September 10 - October 21,1997
-- Inspectors:
S. Ray, Senior Resident inspector P. Krohn, Resident Inspector S. Thomas, Resident inspector Approved by:
J. W. Mc,Cormick Barger, Chief
'
Reactor Projects Branch 7 i
tj 9711'2SO33P 971114 POR ADOCK 05000292
G g gp
,
.
.
-
-.
k EXECUTIVE SUMMARY Prairie Islend fJuclear Generating Plant, Units 1 & 2 NRC Inspection Report No. 50-282/97018(DRP); 50-306/97018(DRP)
This inspection was performed by the resident inspectors and included aspects of licensee operations, maintenance, engineering, and plant support.
Operations Control room operators were attentive to their panels and knowledgeable of plant
conditions and activities in progress. Communications were corwivently clear. Shift turnover briefings were concise but thorough. Pre-job briefing-M ifrequent or complex evolutions were excellent. System engineers were frequently 1 ed regarding their i
systems. No significant problems were noted with routine plan * nerctions (Section O1.1).
Operator control of the plant during two reactor shutdowns and one startup ranged from
good to excellent with no significant problems (Section 01.1).
Operators responded well to an indication of a carbon monoxide buildup in the
control room (Section 01.1).
A conservative operability decision was made regarding the containment maintenance
airloe.k (Section O1.1).
J The Unit 2 safety injection system was properly lined up for safeguards operations with no
significant material condition issues (Section O2.1).
The licensee was making reasonable efforts to resolve the control room habitability
issues in a timey manner (Section O2.2).
The inspectors identified that soveral Section Work Instructions contained operating
instructions that should have had Opt ations Committee reviews but did not (Section O3.1).
Maintenance Procedures were properly used and followed in all maintenance and surveillance activities
observed. Maintenance personnel were experienced and knowledgeable in their tasks.
All activities were performed safely (Section M1.1).
- Good operational control of the p! ant was noted in everal operations surveillance tests (Section M1.1).
Conservative operability decisions were made when on three occasions emergency
diesel generators exhibited abnorms wies during testing (Sections M1.1 and M1.2).
?
,,
-
.
_
....
..-_. _ __. _
_ -. _ _ _ _
_ _. _
_._
Enoineerina j
q Licensee identified findings indicated the engineering staff had conducted a thorough
-.*
review cf logic circuits in accordance'with NRC Generic Letter 96-01. The licensee's -
j-initial actions in response to the findings were conservative (Section E1.1).
~
i
The system engineers were frequently involved in all aspects of plant operations, maintenancef and surveillance activities. The engineers promptly investigated any t
operational abnormalities, took an active role in maintenance and troubleshooting activities, and closely followed all surveillance testing on their systems. The engineers also monitored industry events for applicability to their systems and were frequently observed conducting walkdown inspections (Section E2.1).
The inspectors identified several minor discrepancies in the descriptions of the
.*-
containment pressure instrument penetrations in the Updated Safety Analysis Report (Section E3.1),
Plant Support
- -
The inspectors observed no discrepancies during acquisition and analysis of primary and secondary chemistry samples (Section R1).
Licensee staff performance during the annual medical emergency drill was good
.
(Section P1).-
,
}
.
y q
-- +,. -
-
m
,a
%
-
_
_
-_
__
.
_.
_ _ _ _ _ _
Report Details Summary of Plant Status Unit 1 operated at or near full power until September 29,1997, when a gradual power coastdown was begun. - On October 18,1997, Unit 1 was shutdown from about 82 percent power to begin a scheduled refueling outage. Unit 2 operated at or near full power for the entire inspection period except for a brief forced shutdown on September 26 27,1997 to repair a leaking containment airlock, l. Operations
Conduct of Operations 01.1 General Comments a.
Inspection Scope (71707)'
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of plant operations. These reviews included observations of control room evolutions, shift turnovers, logkeeping, as well as evaluations of operability decisions. Section 13, " Plant Operations," of the Updated Safety Analysis Report (USAR) was reviewed as part of the inspection.
b.
Observations and Findinas
- The inspectors noted that control room operators were attentive to their panels
and knowledgeable of plant conditions and activities in progress, Communications were consistently clear, Shift turnover briefings were concise but thorough. Pre-job briefings for infrequent or complex evolutions were excellent. System engineers were frequently consulted regarding their systems.
No significant problems were noted with routine plant operations; On September 15,1997, a high carbon monoxide alarm occurred on a temporary
portable toxic gas detector in the control room. Operators responded rapidly to
'
the alarm. Actions included verifying the alarm on another portable detector, posting a security guard and maintaining the control room doors open, closing the control room outside air dampers, requesting assistance from the site safety administrator, and inspecting the plant for a possible source of carbon monoxide.
After an alternate sampling technique indicated that there was no carbon monoxide present, the site safety administrator determined that the portable detectors were very sensitive to false indications caused by small amounts of hydrogen gas. Venting of hydrogen to the turbine building roof from the Unit 1 generator was occurring when the alarm occurred. The hydrogen then apparently entered the control room via the open outside control room air dampers. After the hydrogen venting was stopped, the carbon monoxide indication returned to normal.
Hydrogen venting was a relatively common evolution, but this was the first time it was conducted with the outside air dampers open and a portable axic gas
l l
.-.
.
..
-
--.
-
.
. - - ~ ~ -. -
.
.. -
.
-
--
.
i
.
- detector operating in *Se control room. L ~ensee engineers determined that the_
+
concentration of hydrogen was very low and not a habitability or flammability :
j issue,
'
'
q
- eL On September 25,- 1997, normal monthly surveillance procedure (SP) 2132,-
" Personnel and Maintenance Airlock Test," Revision 19, was performed and failed
,
to meet the acceptance criteria.- Licensee engineers attributed the problem to a -
leaking exterior door shaft seal and operators declared the outer door inoperable, The operators entered Technical Specification (TS) 3.6.M.2 limiting condition for -
~,
'
operation which required that the innar airlock door be maintained closed.
On September 26,1997, maintenance workers attempted a repair of the door seal
'
and the airlock was retested. The airlock failed the surveillance a second time.
The licensee entered Unit 2 containment to inspect the inner airlock door and
- noted that the inner airlock door shaft seal was also leaking. This placed Unit 2 in the condition of having two inoperable containment airlock doors.
'
The operators conservatively assumed that both the airlock doors had been leaking during the original surveillance and_ considered the airlock inoperable as of September 25.,1997. Technical Specification 3.6.M.3, for an inoperable airlock, allowed operation to continue for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and then required a unit shutdown.
- Since the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance was almost over by the time the inner door leakage was noted, the operators commenced a Unit 2 shutdown. The licensee notified the NRC, in accordance with 10 CFR 50.72, of the shutdawn. : Unit 2 was taken to hot shutdown in accordance with the TS.
Meanwhile, an emergency work order was written to effect repairs on the airlock.
After the inner and outer airlock door shaft seals were repaired, SP 2132 was repeated satisfactorily.- The Unit 2 maintenance airlock was returned to service and declared operable on September 27,1997. The Unit was restarted later the e
same day.-
The inspectors observed the majority of the reactor shutdown and startup.
Operators properly followed procedures and had good control of the plant. No-significant problems were noted. The licensee intended to issue a Licensee Event Report (LER) for the event. The LER will be considered open, when issued, pending the inspectors' review.
- The inspectors observed the Unit i shutdown for refueling on October 17-18, e
1997. Operator control of the plant was excellent. During the shutdown, licensee personnel performed benchmark testing of the auxiliary feedwater system and '
control rod drop time testing without problems. Historically, the plant experienced control rod deviation alarms during shutdown, so the shift manager assigned the
- Unit 2 shift supervisor to evaluate any deviations to avoid distracting the Unit 1
- operators. Several rod deviation alarms were received. The inspectors had one concem with the supervisor's performance as discussed in Section M1.1 of this report.
_
'
S
-.
.,
.-
.-
.
.-
.,
.
--, -.
. __
_
_
.
.-
.
' During the shutdown, the plant experienced two very brief unavoidable conditions prohibited by TSs and entered the " motherhood" action requirements of
' Specification 3.0.C. The conditions were expected vid the issue was discussed in a letter from the licensee to the NRC, " Technical Specification Interpretation for -
- Auxiliary Feedwater System Operability During Movement into Different Plant Conditions," dated September 15,1997, and in the NRC's response letter,.
" Technical Specification Interpretr,tions for Auxiliary Feedwater and Safety --
' Injection Systems Operability Prairie island Nuclear Generating Plant Unit Nos.1 and 2," dated October 16,1997. The NRC letter indicated that the NRC agreed with some of the licensee's proposed TS interpretations but not with the one involving operability of the auxiliary feedwater pumps with their control switches in
'
the SHUTDOWN AUTO or MANUAL position. The licensee intended to report the condition in an LER and submit a TS amendment request to resolve the operability question. The LER will be considered open, when issued, pending the inspectors' review.
c.
Conclusions Normal operations, as well as two reactor shutdowns and one startup, were conducted in -
a orderly manner with no significant problems. The operators responded well to en indication of a carbon monoxide buildup in the control room. A conservative operability decision was made regarding the containment maintenance airlock.
O2-Operational Status of Facilities and Equipment
- O2.1 Enaineered Safety System Walkdown a.
inspection Scope (71707)
The inspectors performed a walkdown of the Unit 2 safety injection system as part of monthly inspections of engir.eered safety features systems.
b.
Observations and Findinos The inspectors reviewed portions of integrated checklist C1.1.18 2, "SI, CS, CA & HC System Checklist Unit 2," Revision 25, as part of the Unit 2 safety injection system walkdown. Local valve positions, tags, blocks, caps, and switch positions, as well as control room indications and switch positions, were verified to be in the proper safeguards positions. The inspectors observed the material condition of the system and noted no significant discrepancies, c.
Conhslons
- The inspectors found the Unit 2 safety injection system to be properly lined up for safeguards operations with no significant material condition issues.
-
.g._y
.-.
.
-.
.
-...
-
. -.
-
.-
i O2.2 - Status of Control Room Special Ventilation System and Habitability
[nfoection Scope (92901)
a.
n At the conclusion of the last inspection period, the licensee was maintaining the control room outside air dampers closed pending resolution of concerns relating to various control room habitability issues. These issues were previously documented in inspection Report 50 282/97015(DRP); 50-306/97015(DRP), Section E2.2, and were considered an Unresolved item (URI 50-282/97015-04(DRP); 50-306/97015-04(DRP)). In the cover letter for the inspection report the NRC requested that appropriate attention be given to resolving the issues in a timely manner, During this inspection period, the licensee engineering and operations staffs resolved the issues to the point that they decided to reopen the dampers, The inspectors met with the licensee staff, including the general superintendent engineering (GSE) on October 17,1997, to discuss the status of the resolutions.
b.
Observations and Findinas The GSE informed the inspectors that initial calculations regarding the effect of a carbon dioxide fire suppression actuation in the relay room, with the relay room / control room dampers failing to close, indicated that carbon dioxide levels in the control room would not approach hazardous levels. The calculations were being finalizea by a vendor and were expected to be completed by the end of 1997, In the meantime, the relay room / control room dampers were being maintained closed.
The GSE informed the inspectors that initial calculations regarding the onsite rupture of a carbon dioxide tank, hydrazMe tank, or propane tank showed that the rupture would not adversely affect the control room. The calculations were being finalized by a vendor and were expected to be completed shortly. The GSE stated that he had a Management Action Tracking item assigned to himself to add a program to Section H of the Operations Manual to insure that hazardous substances stored onsite, which exv.aded the amounts in the Superfund Amendments and Reauthorization Act, would be evaluated for possible effe on control room habitability.
The GSE informed the inspectors that the statement in the NRC letter of March 4,1985, stating that the NRC required institution of the training program the licensee proposed to assure toxic gas detection by the operators, went beyond the training intended in the licensee's proposal as documented in their letter of July 9,1984. In that letter, the licensee committed to human detection along with administrative controls and training for control room personnel to provide protection from the toxic gases. The GSE indicated that training in the detection of actual toxic gases would create a health concern and that they would institute the commitment change process to clarify the commitment intent.
The GSE indicated that the odor identification test that all licensed operators were given in their biannual medical examination was sufficient to insure that human detection could be accomplished.
The GSE informed the inspectors that operators received annual classroom training in donning self contained breathing apparatus (SCBAs). Operators could choose to actually
~ don the SCBAs during that training, if desired. The inspectors expressed the concern
'
that the training did not insure that operators practiced actually donning the SCBAs. The
.
.
_
_
_
-.
..
-
.
-_
-
.
,
- inspectors' discussions with operators indicated that, except for fire brigade members, few operators had actualh donned an SCBA for several years. The inspectors verified.
,
'that the licensee staged two additional sets of SCBAs in the control room, for a total of six. Additional sets of SCBAs were maintained in the adjacent records room. The inspectors verified that SCBA compatible corrective lenses for operators required to wear them were being placed in close proximity to the control room but the action was not
?
. complete, Additional concerns with control room habitability and the control room special ventilation system were documented in inspection Report 50-282/97008(DRS); 50-306/97008(DRS),
Sections E1,4, E1.5, and E3.2. One of those concerns was considered a URI (50- 282/97008 09(DRS); 50-306/97008-09(DRS)). The GSE informed the inspectors that a control room ventilation envelope inleakage test was being developed and was planned to be run in January 1998. The licensee intended to complete new control room dose calculations based on the results of the testing. A modification to install a permanent safety-related alternate air system for the special ventilation system's control valves was being planned for installation in November 1997. Additional modifications were being developed to upgrade the system based on the Seismic Qutification Users Group methodology.
Licensee Event Report 197 06 discussed an additionalissue with the system conceming lack c';sdundancy of radiation monitors. The GSE informed the inspectors that a modification to resolve that issue was in development and was planned for installation in early 1998. In the meantime, control room outside air dampers were being maintained closed during fuel handling operations, c.
Conclusions The inspectors concluded that the licensee was making reasonaLie efforts to resolve the control room habitability issues in a timely manner. The URIs and LER discussed above remain open, pending completion of the actions discussed.
O3 Operations Procedures and Documentation O3.1 Review of Operations Section Work Instructions inspection Scope (92901)
a.
n in response to a finding by NRC operator license examiners, the inspectors reviewed the.
operations department Section Work Instructions (SWis) to determine if TS review requirements were met.
b.
Observations and Findinas The licensee maintained a series of SWis to provide guidance to operations personnel.
Most of the SWis were administrative in nature and provided instructions for such things -
as general watchstanding protocol, logkeeping, and duties of various positions. However, the inspectors determined that some of the SWis contained operating instructions of a l
i
!
-
_.
_
.
.
_
,
_
-.
_
! nature that they fit into' the requirements of TS 6.5 for phnt operating procedures
_
requiring review by the Operations Committee (OC). i ne of the operations SW!s had received OC reviews; The inspectors determined that some of the SWI operating instructions were merely duplications of instructions contained in other procedures that were OC reviewed.
However, some of the instructions were not included in any other procedures. The -
inspectors turned over several examples of SWI instructions that appeared to need OC review to the operator license examiners for inclusion in their report (50 252/97019(DRS); 50-386/97019(DRS)).
c.
Conclusions-
- The inspectors concluded that the licensee had several SWis that contained operating instructions that should have had OC reviews but did not.
II. Maintenance M1 Conduct of Maintenance M1.1 General Comments a.
Inspection Scope (61726. 62707. 92902)
The inspectors observed all or portions of the following maint3 nance and surveillance activities, included in the inspection was a review of the surveillance procedures (SP),
preventative maintenance (PM) procedures, or work crders (WO) listed as well as the appropriate USAR sections regarding the activities. The in:;pectors verified that the surveillance procedures observed met the requirements of the Technical Specifications.
- SP 1089 Residual Heat Removal Pumps and Suction Valves From the Refueling Water Storage Tank, Revision 46
SP 1106B 22 Diesel Cooling Water Pump Test, Revision 52
~
e SP 1130 Containment Vacuum Breakers Guarterly Tests, Revision 28
SP 1218 Monthly 4KV Bus 15 Undervol; age Relay Test, Revision 21 e
SP 1258A Bus 15 Sequencer Load Rejection and Restoration of 1:1 Control Room Chiller, Revision 0
SP 1258B Bus 16 Sequencer Load Rejection and Restoration of 122 Control Room Chiller, Revision 0 e
SP 1319 Rod Position Verification, Revision 4
SP 1669 Computer Out of Service Log - Unit 1, Revision 19
'
o SP 2093 D5 Diesel Generator Slow Start Test, Revision 64
,
SP 2130 Containment Vacuum Breakers Quarterly Tests, Revision 33 e
SP 2307 D6 Diesel Generator Fast Start Test, Revision 10 e
- SP 2335
. D6 Diesel Generator 24 Hour Load Test, Revision 7
SP 2102 22 Turbine-Driven Auxiliary Feedwater Pump Test, Revision 51
WO 9706898 - PM 3120-52-1, Main Steam Safety Valve Test (Hot), Revision 4
- -
WO 9707142 PM 3124-1-12,12 RHR Pump Annualinspection
WO 9702548 Remove / Replace Snubber 1-cch-398b o'
WO 9702546 Remove / Replace Snubber 1-cch-398a
4
,e
~
v
m WO 9702545 Remove / Replace Snubber 1-cch-397
- -
. WO 9706374 D6 Emergency Diesel Generator 18 Month Preventative
- '
Maintenance
- *-
WO 9708995 Move Fuelin Spent Fuel Pool to Checkerboard
-*-
WO 9711480 l Rod Drop Test
' WO 9711814 Benchmark Testing of the 12 Auxiliary Feedwater Pump b.
Observations and Findinas
. For all of the work observed, procedures were properly used and followed, Maintenance -
personnel were experienced and knowledgeable of their tasks. The inspectors observed frequent monitoring of work by system engineers.
_
Tne inspectors observed good control by the Unit i reactor operator during
performance of SP 1089, " Residual Heat Removal Pumps and Suction Valves From the Refueling Water Storage Tank." Operations personnel were stationed in the 12 residual heat removal pump enclosure during the test. High noise levels
-
and surrounding concrete enclosures made radio communications with those operators difficult. When confusion arose, the reactor operator stopped the performance of SP 1089. Operations personnel relocated to a quieter area, re-established an understanding of the steps being performed, and completed the test satisfactorily.
Inspectors observed good control during the performance of SP 1130,
" Containment Vacuum Breakers Quarterly Tests." Operators were stationed in the control room, annulus area, and locally at the differential pressure switches that were being tested by the survulance. Coordination and communications between all stations was good.
During the pre-evolution brief, the inspectors noted that SP 1130 listed an incorrect drawing, not applicable to the system being tested, in its list of references. The shift supervisor was informed and, since the error did not impact the performance of the SP, a correction was made to the working copy and a procedure revision request was submitted.
The inspectors observed portions of WO 9706374, "D6 Emergency Diesel
Generator 18 Month Preventative Maintenance? The inspectors witnessed removal and installation of various instrumentation, sensors, and fuel injectors as well as valve lash and clearance adjustments. On September 19,1997, during a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-maintenance test 100 percent load run, the operators observed that one of the cylinders (cylinder B1 on engine 2) had an exhaust temperature approximately 150 degrees Fahrenheit higher than the average exhaust temperature of the other cylinders. The operators conservatively shut down the -
engine, declared the engine inoperable, and had maintenance workers recheck the injector and valve adjustments for that cylinder. The post-maintenance testing was successfully repeated on the afternoon of September 19,1997.
During the retest however, the engine experienced intermittent alarming of the exhaust gas high temperature alarm. Troubleshooting revealed the cause to be a malfunctioning temperature scanner module located at the local diesel operating
station console. The scanner module provided only localindication and alarm signals and was replaced. The inspectors witnessed the fast start testing of D6 emergency diesel generator and noted that the engine achieved rated fre aency in voltage in 7 seconds, meeting the requirement of 10 seconds.
s Inspectors observed spent fuel handling operations per WO 9708995, " Move Fuel
in Spent Fuel Pool to Checkerboard." Good control of the evolution was observed, especially in the aspect of verification. Prior to grappling a fuel assembly and prior to placing a fuel assembly into the spent fuel pool, three-party verification occurred.
Each location in the spent fuel pool was designated by a number and a letter.
The numbered rows ran east and west, the lettered rows ran north and south. To locate the correct pool location, the bridge crane operator moved the bridge to the correct numbered row, as indicated by numbers on the north and south sides of the spent fuel pool and then positioned the bridge crane to the correct letter row, as indicated by an indexing system located on the movable bridge rail. During fuel handling operit sns, the inspectors observed that there were also letters painted on the wall at tue east end of the spent fuel pool which did not agree with the letter indexing system located on the movable bridge. When questioned about the inconsistency, the nuclear engineer stated that the letters on the wall were probably obsolete references from before the spent fuel pool was reracked. He indicated that those letters were not used as a reference to position the bridge and that action would be taken to correct the inconsistency. While observing fuel handling operations the following day, the inspectors noted that the letters located on the east wall of the spent fuel pool had been covered and were no longer visible.
inspectors observed the pre-evolution briefing for and performance of SP 2102 e
"22 Turbine-Driven Auxiliary Feedwater Pump Test." The conduct of the brief ey and the test were performed in a satisfactory manner.
In past operations of the 22 turbine-driven auxiliary feedwater pump, it was observed that a pressure transient occurred in the suction line of the pump, subsequent to the shutdown of the pump. On at least one occasion, the suction pressure gauge associated with this pump was over-ranged as a direct result of the pressure transient and had to be replaced. The full-scale indication for the gauge was 15 pounds per square inch - gauge (psig). To prevent damaging the suction pressure gauge during the performance of the procedure, special test equipment was installed which allowed for isolating the pressure gauge and recording the pressure transient during pump shutdown. At the completion of the procedure when the pump was secured, a pressure transient lasting approximately 3 seconds, with a peak pressure of approximately 98 psig, occurred.
One factor that was not considered, until questioned by the inspectors, was the impact of the pressure transient on the 22 turbine-driven auxiliary feedwater pump's low suction pressure trip svvitch. That pressure switch was installed on the same sensing line as the suction pressure gauge. The low suction pressure trip switch was designed to trip at 4 inches of mercury pressure, to be able to
function with sustained system pressure of 50 poig, and to be able to withstand a brief pressure transient up to 250 psig As a result of the inspectors' concern, the licensee evaluated the pressure switch for proper operation. When tested, the switch was found to be functioning properly but was out of calibration. The pressure switch operated at 3.4 inches of mercury, with the required setpoint being 4 inches +0.1 inch. Licensee instrument technicians then recalibrated the pressure switch setpoint and the pump was declared to be operable.
A preliminary evaluation by the licensee to discover the cause of the suction line pressure transient determined that it was leakage past the auxiliary feed pump's discharge check valve As a temporary corrective action until the check valve could be repaired, the licensee issued Temporary Memo TMA-1997-0150 which modified the 22 turbine-driven auxiliary feedwatar pump operating procedures to instruct operators to isolate the suction pressure gauge and low suction pressure trip sensing line just prior to pump shutdown and to return it to service immediately after the pump is secured. Additionally, licensee engineers evaluated whether the "as found" pressure switch setpoint impacted accident analysis and/or placed the system in a condition outside of the design basis. Their conclusion was that operability was not affected. The engineers determined that the switch had a history of setpoint drift and were considering permanent corrective actions for that issue.
inspectors observed testing of the Unit 1 loop A main steam safety valves on e
October 18,1997. All five loop A valves met the "as found" acceptance criteria of +/ 3 percent f am the specified setpoint value. Two of the five safety valves had to be adssted so that the "as left" criteria of +/-1 percent was met.
Dt ing the shutdown of Unit 1 for refueling, several rod deviation alarms were e
received on control bank D rods. Because such alarms were common during large power changes, the shift manager assigned the Unit 2 shift supervisor perform SP 1319, " Rod Position Verification," to avoid distractions for the operators conducting the shutdown. A rod deviation was first received for rod G-11 and the supervisor started the surveillance procedure. Step 7.1.6 of SP 1319 required the operator to obtain a thermocouple map after the rod deviation, which he did. The map was used to verify that the control rod was not actually misaligned by comparing core exit thermocouple readings adjacent to the affected rod to those in symmetric core locations.
About 2 minutes after obtaining the thermocouple map, the licensee received two additional rod deviation alarms for rods C-7 and K-7. The supervisor added those rods to the surveillance. However, the inspectors noted that he did not obtain another thermocouple map after the additional deviations and used the original map to venfy that rods C-7 and K-7 were not misaligned. When he had completed the calculations to his satisfaction, the inspectors pointed out the observation and the supervisor obtained another thermocouple map and reperformed the verification. The calculations demonstrated that none of the rods were actually misaligned.
The inspectors discussed the observation with the general superintendent of operations (GSO) who investigated the issue further. The GSO determined that at the time the first thermocouple map was obtained, rods C-7 and K-7 were already deviating from the group position, but had not yet reached the alarm setpoint of greater than 12 steps. The rods were only moved one additional step from the time of the first thermocouple map until the alarms were obtained. The GSO stated that the supervisor had met the procedure requirement of obtaining a thermocouple map after the deviation (the procedure did not say to obtain the map after the alarm).
The inspectors agreed that the procedure had not be,n violated. Hovvever, the inspectors pointed out that the purpose SP 1319 was to verify that control rods were not actually misaligned. Technical Specification 3.10.E.2.b required that, if the bank demand position was between 30 and 215 steps and the rod position channel differs by more than 12 steps, that rod cluster control assembly be considered misaligned. Thus, using SP 1319 to show that rods indicating less than a 12-step deviation were not actually misaligned, would not be adequate to show that the rods were still not misaligned after they reached the TS of more than 12 steps.
In this case, the rod indications all retumed to within 12 steps of the group positions well before the 8-hour time allowance in Technical Specification 3.10.E.3. Therefore, successful completion of SP 1319 was not necessary to meet TS requirements. Thus, no violation of NRC requirements occurred.
c.
Conclusions During this inspection period, the inspectors observed a wide variety of maintenance and surveillance activities, most of which were performad well. The inspectors pointed out a few minor problems which were adequately addm ;ed by licensee personnel. All activities were performed safely.
M2.1 Emeraency Diesel Generator Oil System Leaks a.
Inspection Scope (37551)
The inspectors observed a 24-hour load run of the D1 emergency diesel generator.
During the load run the engine developed a lubricating oilleak on an elbow between the lubricating oil pump discharge and the oil filter housing. Licensee engineers contacted the vendor and the engine owner's group and learned that the similar leaks had occurred in the same location at other nuclear facilities. The inspectors observed the repairs and subsequent licensee actions to address the generic issue.
!
Observations and Findinas The licensee discovered the pinhole leak on September 30,1997. Operators secured the D1 emergency diesel generator and entered a 7-day allowed ou'c je time in accordance with TS 3.7 B.1. The TS required the operability of D2 emergency diesel generator to be demonstrated within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
_
-.
_
~ _
- -.
- -
-
. - -. _ -
Ai d
While D2 operability was b ng emonstrated later the same morning, an oil system leak;
- on a lubricating oil supply line to the turbocharger developed < The licensee secured the engine and declared D2 inoperable. That placed Unit 1 in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowed outage time in accordance with TS 3.7,B.5 after which they were required to initiate action within one 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to conduct a shutdown in becordance with TS 3.0.C. While repairing the D2 -
turbocharger oil system leak, the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time limit was exceeded at 1139 a.m. The licensee notified the NRC i_n accordance with 10 CFR 50.72 for commencing a shutdown
- required by TSs.' The oilleak on the turbocharger was repaired and D2 returned to operable status at 12:12 p.m. on September 30,1997, before power was reduced.
The licensee removed the leaking lubricating oil pipe from D1 and performed an inspection to detumine the cause of the pinhole leak. The licensee concluded that the
'
leak was caused by an incomplete penetration weld This type of failure had been seen at several other facilities with the same type of engine. Weld materialin the vicinity of the
' pinhole leak was removed and the elbow repaired with a full penetration weld. As a precaution, the identical weld on D2 was inspected. No discrepancies were noted. The vendor noFfied the NRC and potentially affected plants of the generic issue in a letter
. dated September 30,1997, c.
ftpnghsions
.
Operators made conservative operability decisions and properly followed the TSs. The Di piping was properly repaired and inspected. D2 piping was appropriately inspected as a precautionary measure, Licensee engineers consulted with other industry personnel and promptly reported their findings to the vendor. As a result of the licensee's contact with the vendor, the vendor made a notification to the NRC and other potentially affected plants, lil, Enaineerina E1 Condect of Engineering E1.1 Coolina Water Pumo Outside of the Desian Basis a.
Inspection Scope (92903).
On October 3,1997, the licensee reported in accordance with 10 CFR 50.72 that the
'
'121 motor-driven cooling water pump was in a condition outside of the design basis. The inspectors reviewed the circumstances of the finding and the initial actions taken by the.
- licensee.
b.
Observations and Findinas
~ During a review required by NRC Generic Letter 96-01, " Testing of Safety Releted Logic Circuits," the licensee identified two issues. The first was that the automatic start of the 121 cooling water pump on low cooling water header pressure was not being tested. The low pressure start was relied on to mitigate the consequences of a loss of offsite power event when the 121 cooling water pump was being relied on as a replacement for either of the safety related diesel-driven cooling water pumps. The second issue was that the pressure switch for the pump was located in the same panel
,
a.
.
..
6-
,
.r.,,.
,,
,.
.
y p
,
y..
...,. -
as the similar switch for the 12 diesel-driven cooling water pump and was not electrically separated by at least 36 inches as required by the USAR. Thus, if the 121 pump was being relied on as a replacement for the 22 diesel driven cooling water pump, a single failure in the panel could possibly disable the low pressure start capability of both the 12 and 121 safety-related pumps.
At the time of discovery, the 121 pump was not being used as a replacement for either of the ditsel-driven pumps. The licensee declared the 121 pump inoperable and started developing a test procedure for the low pressure start and a modification to separate the pressure switches. The licensee had high confidence that the low pressure start feature woulf, have operated properly because the pressure switch itself was in a regular calib ation program, and the pum,n had actualiy started on low pressure several times in the eecent past due to various events effecting cooling water header pressure. The lice 1see intended to issue an LER on the findings. The enforcement aspects of the findings will be reviewed when the LER is issued.
c.
Conclusions The licensee-identified findings indicated a thorough review of logic circuit testing was being accomplished in accordance with NRC Generic Letter 96-01. The licensee's initial actions in response to the findings were conservative.
E2 Engineering Support of Facilities and Equipment E2.1 General Comments Throughout the inspection period the inspectors noted frequent involvement by system engineers in all aspects of plant operations, maintenance, and surveillance activities.
The engineers promptly investigated any operational abnormalities, took an active role in maintenance and troubleshooting activities, and closely followed all surveillance testing on their systems. The engineers monitored industry events for appdcability to their systems and were frequently observed conducting walkdown inspections.
E3 Engineering Procedures and Documentation E3.1 USAR Discrepancies a.
Inspection Scope (92903)
While reviewing a surveillance activity (SP 2130, Revision 33), the inspectors noted that the actus' configuration of the containment penetrations for containment pressure instrumentac in did not match the USAR. The inspectors reviewed the following documents:
USAR Section 5.2.2.1.1, " Isolation System," Revision 14
USAR Table 5.2-1-(Part A), " Unit 1 Containment Vessel Penetrations," Revision 5
USAR Table 5.2-1-(Part B), " Unit 2 Containment Vessel Penetrations," Revision 5
USAR Figure 6.31, Revision 14 (Drawing NF-39602-1, " Reactor Building Unit 1
Ventilation Flow Diagram," Revision AH)
-.
.
-
..~ -
-.
-
-
Le Drawing NF 39602 2, * Reactor Building Unit 2 Ventilation Flow Diagram,"
-
Revision AK.
- USAR Section 7.4.2.2.1, " Containment Pressure," Revision 14 -
- -
USAR Table 7.4 4," Process Instrumentation for RPS (reactor prctection system]
& ESF [ engineered safety features] Actuation," Revision 9 USAR Table 7.4.5, " Post Accident Equipment (Inside Containment) Operability e
- Requirements," Revision 0 USAR Table G.2 2, *Through-Line Leakage that Bypass Annulus and Terminates __
e=
in ABSVZ [ auxiliary building special ventilation zone)," Revision 4 USAR Table G.2 3,"Through Line Leakage that Bypasses Both Annulus and e
ABSVZ," Revision 13 b.
Observations and Findinal The inspectors noted the following USAR discrepancies:
USAR Table 5.21 listed only two penetrations as containing instrumentation lines:
36D and 49A. Actually, a review by the inspectors of drawings NF-396021 and NF-39602-2, as well as a field walkdown, showed that four penetrations:
3,360,42A, and 49A on Unit 1, and 3,42A,49A, and 56 on Unit 2, contained containment pressure instrumentation lines. The inspectors also noted that penetration 42A contained a containment pressure line in the shield building penetration but the line was routed through penetration 50 in the containment.
The main purpose of the table was to describe the isolation valves and testing requirements for the various penetrations. Since the instrument lines had no containment isolation valves nor testing requirements beyond the integrated leak rate test, the missing information on the table was not significant.
e Penetration 56 on Unit 2 was not listed in Table 5.21. The inspectors were
<
informed by an engineering superintendent that it may have been mislabeled and might actually be penetration 36D as in Unit 1.
e Table 5.21 listed penetration classes for each penetration as defined in USAR Section 5.2.2.1.1, The table listed penetration 36D as Class 2 (not required to be open for assumed post accident conditions). Actually, the line supplied instrumentation for actuation of safety injection, containment spray, arJ containment vacuum breakers, all required for post-accident indication and mitigation as discussed in USAR Section 7.4.2.2.1 and listed in USARTable*r.4 5.
It possibly should have been listed as Class 6 (systems reqdred to operate e the post-accident condition). The various classes primarily described isolation i cWe configurations. The instrument lines had no isolation valves so it was not really significant how they were classified.
Table 5.21 listed penetration 49A as Class 5 (normally closed to the reactor
vessel atmosphere during reactor operation). Actually the cor.tainment pressure
_
monitoring lines are continuously open to the atmosphere inside containment and possibly should have been listed as Class 6. The engineering superintendent-stated that the lines are closed outside of containment and Class 5 might be appropriate. Although each containment penetration was categorized into one of
- the classes depenoing on isolation valve configuration, the inspectors could find -
.
-
.
.-
-
-
no analysis which made use of that information. Therefore, an improper classification was not considered a significant issue.
USAR Table G.2 3 lists penetrations 3,36D,49A, and 50 as instrumentation lines
from which leakage would bypass both the annulus and the ABSVZ. Actually, all portions of the lines outside of the shield building were completely inside of the ABSVZ. Those lines possibly should have been listed in Table G.2-2 as lines
,
from which leakage would bypass the annulus and terminate in the ABSVZ.
However, all entries in Table G.2 2 had been deleted from the USAR in an earlier revision. Assuming the leakage from the instrument lines would bypass the ABSVZ appeared to be a conservative assumption for release calculations.
Therefore, the issue was not considered significant.
- USAR Section 7.4.2.2.1 discussed six containment pressure channels organized in three pairs of transmitters with each pair having its own pressure tap connection to the containment. Actually the six channels were arranged in two pairs of transmitters and two single transmitters for a total of four pressure tap connections to the containment.
Similarly to the above, USAR Table 7.4-4 listed six containment pressure
transmitters with three shared taps. Actually there were four taps for the six transmitters.
10 CFR 50.9(a) required, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. The information provided by the licensee regarding the configuration of the containment pressure instrumentation penetrations in the USAR was not accurate as described above.
The licensee had recently initiated a comprehensive USAR review and update project, which was described in a letter to ths NRC dated September 26,1997. The licensee project manager also discussed the 6 roject and recent findings with the inspectors during this inspection. The inspectors concladed that the project, as described, would have eventually found the discrepancies. Tha discrepancies were of minor safety significance.
Having the containment pressure instrumentation distributed over a larger number of penetrations than the USAR described made it less likely that the protective function would fail. Assuming that leakage from the lines would bypass the ABSVZ made the offsite release calculations more conservative. In addition, plant drawings contained accurate information regarding the containment pressure instrumentation penetrations.
Thus, this failure constitutes a violation of minor significance and is being treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforcement Policy (50-282/97018-01(DRP); 50-306/97018-01(DRP)).
c.
Conclusions The inspectors concluded that the USAR discrepancies were of minor safety significance and would eventually have been discovered by the licensee's USAR review project. The licensee added the discrepancies to their corrective action program.
IV. Plant Support R1 Radiological Protection and Chemistry Controls (71750)
During normal resident inspection activities, routine observations were conducted in the areas of radiological protection and chemistry controls using inspection Procedure 71750. The inspectors accompanied the licensee during acquisition and analysis of a full set of primary and secondary samples on September 18,1997. Procedures were followed adequately. The post accident chemistry monitoring system was also reviewed. No discrepancies were noted.
P1 Conduct of Emergency Preparedness Activities (71750)
During normal resident inspection activities, routine observations were conducted in the area of emergency preparedness using Inspection Procedure 71750. The licensee conducted an annual medical emergency drill on September 16,1997. Federal Emergency Management Agency personnel were present to monitor the offsite portion of the exercise. Local fire, ambulance, and hospital personnel and facilities participated in the contaminated injured man drill. The inspectors observed the onsite portion of the drill. Drill simulation and control were good. The first responders to the scene of the contaminated injured man properly administered to the medical concerns while taking appropriate radiological precautions. No instances of uncontrolled spreading of radioactive contamination occurred for the portions of the drill observed onsite.
S1 Conduct of Security and Safeguards Activities (71750)
During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities using Inspection Procedure 71750. On September 17,1997, several Federal Bureau of Investigation investigators and Directors and local law enforcement officials came to Prairie Island for discussions concerning potential terrorist threats and site familiarization tours. The inspectors participated in the dialogue and established contacts with the officials. No discrepancies were noted in the licensee's security postures or plans.
F1 Conduct of Fire Protection Activities (71750)
The inspectors examined licensee procedures and methods for controlling transient combustibles in the turbine and auxiliary buildings. Unit 1 entered a refueling outage during the inspection period which resulted in a large amount of wood and other transient combustibles being introduced into the site. The inspectors checked the licensee fire hazard analysis and noted that the limits on transient combustibles were not exceeded.
F2 Status of Fire Protection Facilities and Equipment a.
Inspection Scope (71750)
The inspectors reviewed emergency diesel generator remote tachometer indication circuitry design to verify that a control room fire causing a hot short affecting the tachometer would not render inoperable other emergency diesel generator protection or control functions. The inspectors specifically reviewed the circuitry for the D1 and D5 diesels because the licensee had designated those diesels as the 10 CFR 50, Appendix R, safe shutdown diesels for shutdown from outside of the control room.
!
!
b.
Observations and Findinas The inspectors reviewed the following D1 emergency diesel generator remote tachometer circuitry design drawings:
e NE-40009, Sheet 72, Revision DC e
NE 40009, Sheet 73, Revision DC e
NE-40009, Sheet 74, Revision DY
NE-40009, Sheet 75, Revision DX e
NE-40009, Sheet 79, Revision AG The inspectors found that for D1, local isolation control switch CS 55413 isolated the remote emergency diesel generator tachometer circuitry in the control room when taken from the * remote" to the " local" position. The inspectors verified that the licensee procedure for shutdown from outside of the control room instructed the operators to place that switch in " local."
The inspectors reviewed the following D5 emergency diesel generator remote tachometer circuitry design drawings:
NE 116756, Sheet 7, Revision A e
NE-116756, Sheet 8, Revision A
NE-116756, Sheet 11, Revision A e
NE 116756, Sheet 28, Revision B e
NE-116756, Sheet 33 Revision A The inspectors found that there were six speed sensors on each of the D5 engines (D5 had two separate diesel engines driving one electrical generator). One of the speed sensors on each engine supplied an independent remote tachometer indication located in the control room. All of the generator's spaed control and protective functions were supplied by other speed sensors. Since the remote tachometer circuit was independent of other emergency diesel generator control and protective functions, a hot short in that circuit would not affect local control of the D5 generator, c.
Conclusions The inspectors concluded that a control room fire damaging remote diesel generator tachometer circuitry would have no effect on the operability of the D1 and D5 emergency diesel generators from local control stations.
V. Manaaement Meetinos X1 Exit Meeting Summary The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on September 21,1997. The licensee acknowledged the findings presented. At the exit meeting, the licensee provided additional information regarding some of the inspectors' preliminary findings. That information was incorporated into this report where appropriate. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
-.
.
- - _. - _.
- _..
~.._ _.... _ _._.- _
- __. __.._..._-_.... __.__...- __._
.
i
. =
!
PARTIAL LIST OF PERSONS CONTACTED
'
Licensee
J.!Sorensen, Plant Mar.ager -
K. Albrecht, General Superintendent Engineering ElectficaPinstrumentation & Controls LT, Amundson, General Superintendent Engineering, Mechanical.
i
- - -
J.' Goldsmith, General Superintendent Engineering, Generation Services J. Hill, Manager Quality Services
LG, Lenertz, General Superintendent Plant Maintenance -
-J; Maki, Outage Manager
.
_
.
D. Schuelke; General Superintendent Radiation Protection and Chemistry T. Silverberg, General Superintendent Plant Operations M. Sleigh, Superintendent Security -
,
_
o
- 20 c.
-
w p--
ws.
ve---
+
, - =
S,E-w-
gyearsgy-e.,y y-giy-gm--r 1--r q
q y-m-=
yms
--i---<-
r-
---yy-
'+g.-=v--m.,
y w
w
w
,,. raw r
.
- -.
.
_.
.. -..
. -. - -
.. - - - - - - - -
.-
.-
INSPECTION PROCEDURES USED
, IP 37551:
- Engineering
' IP 61726:
Surveillance Observations J
-
,
- IP 62707:
Maintenance Observations
. IP 71707:
Plant Operations
,
IP 71750:
Plant Support Activities
_
+
iP 92700:
Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor -
Facilities IP 92901 Follow up - Operations IP 92902:
Follow up - Maintenance IP 92903:
Follow up Engineering IP 97.904:
Follow up - Plant Support IP 93702:
Prompt Onsite Follow up of Events ITEMS OPENED, CLOSED, AND DISCUSSED
- Opened 50-282/97018-01(DRP)
NCV Discrepancies in the SAR Regarding 50-301/97018 01(DRP)
Containment Pressure instrumentation Penetrations Closed None.
Discussed 50-282/97015 04(DRP)-
URI - Control Room Habitability 50-306/97015-04(DRP)
50-282/97008-09(DRS)
URI Control Room Special Ventilation System 50-306/97015-09(DRS)
50-282/07006 LER Discovery of Logic Error in Control Room Ventilation System l
l-p
21 I
.
.
-
_
--.
--
.....
..
-
.
. ~..
-
-.
.
.
.
t LIST OF ACRONYMS USED.
,
ABSVZ_
. Auxiliary Building Special Ventilation Zone
-
CFR
- Code of Federal Regulations DRP:
Division of Reactor Projects.-
DRS'
Division of Reactor Safety GSE General Superintendent Engineering GSO.
General Superintendent Operations -
.IP inspection Procedure LER Ucensee Event Report ~
NCV-Non-cited Violation-NRC'
_ Nuclear Regulatory Commission OC.
Operations Committee-PDR Public Document Roorn
. psig Pounds per Square Inch Gauge PM Preventive Maintenanca
.SCBA.
Self Contained Breathing Apparatus
- SP Surveillance Procedure -
' SW1 Section Work Instruction -
=TS Technical Specification USAR Updated Safety Analysis Report
- WO.
Work Order -
. -.
--
.
._
-
.