IR 05000285/1991014
| ML20024H549 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/23/1991 |
| From: | Ray Azua NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20024H547 | List: |
| References | |
| 50-285-91-14, NUDOCS 9106050006 | |
| Download: ML20024H549 (15) | |
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APPENDIX
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-285/91-14 Operating License:
DPR-40 Docket:
50-285 Licensee: Omaha Public Pcwer District (OPPD)
444 South 16th Street Mall Omaha, Nebraska 63102-2247 Facility Name:
Fort Calhoun Station (FCS)
Inspection At:
FCS, Blair, Nebraska inspection Conducted: April 10 through May 17, 1991 Inspectors:
R. Mullikin, Senior Resident Inspector T. Reis, Resident Inspector
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Insyection Summary Inspection Conducted April 10 through May 17, 1991 (Report 50-285/91-14)
Areas Inspected:
Rout.ne, t... announced inspection of the review of a previously identified inspection finding, review of corrective actions for violations, licensee event report (LER) followup, onsite followup of events, operational safety verification, and maintenance and surveillance observations.
Results:
Apparent personnel errors resulted in the loss of power to the technical support center and initiation of a ventilation isolation actuation signal (paragraphs 6.a and 6.c).
The plant operators promptly and effectively responded to a plant perturbation when a main feedwater regulating valve began oscillating (paragraph 6.b).
Control room activities were found to be properly performed (paragraph 7.a).
Radiation protection and security personnel were observed performing their duties in an effective manner (paragraphs 7.c and 7.d).
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a-Licensee quality assurance audit and surveillances and resulting
< corrective actions were found to be comprehensive and prompt (paragraph 7 f).
- Maintenance activities were found to be performed in accordance with procedures, Good coordination was noted between maintenance and quality control personnel (paragraph 8).
Surveillance activities were observed being performed according to procedures with good attention to detail (paragraph 9).
Note:
Acronyms and initials used in this report are identified, in an alphabetical listing, on an attachment at the end of this inspection report.
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DETAILS 1.
Persons Contacted
- R, Andrews, Division Manager, Nuclear Services T. Dailey, System Engineer
- S. Gambhir, Division Manager, Production Engineering
- J. Gasper, Manager, Training
- W. Gates, Division Manager, Nuclear Operations
- R. Jaworski, Manager, Station Engineering
- L. Kusek, Manager, Nuclear Safety Review Group
- T. Matthews, Acting Supervisor, Station Licensing
- W. Orr, Manager, Quality Assurance and Quality Control
- T. Patterson, Manager, Fort Calhoun Station
- R. Phelps, Manager, Design Engineering A. Richard, Assistant Manager, Fort Calhoun Station C. Schaffer, System Engineer
- J. Sefick, Manager, Security Services C. Simmons, Station Licensing Engineer F. Smith, Supervisor, Chemistry
- T. Therkildsen, Acting Manager, Nuclear Licensing and Industry Af f airs D. Trausch, Supervisor, Operations Tha inspectors also contacted additional personnel during this inspection period.
"Danctes attendance at the monthly exit interview held on May 17, 1991.
2.
Plant S atus The FCS operated at 70 percent power from the beginning of this inspection period until April 23, 1991, when a reduction in power to 35 percent was commenced.
The power reduction was necessary to correct oscillations in a main feedwater regulating valve (FCV-1101).
The licensee isolated the air supply to the valve, secured Valve FCV-1101 at the 44 percent open position, and maintained steam generator level by use of a feedwater bypass valve (HCV-1105).
The licensee made the necessary repairs and began power ascension on April 27.
The plant reached 75 percent power on April 29, where it remained throughout the rest of this inspection period.
3.
Review of a Previously_ Identified Inspection Findirig__(_92701]
(Closed) Unresolved Item 235/B928-01:
Potentially inadequate engineering evaluation.
During the summers of 1989 and 1990, it was discovered that high ambiert temperatures in the emargency diesel generator (EDG) rooms caused the EDGs to operate at elevated temperatures.
The licensee noted that the EDG i
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cooling systems did not function properly.
In addition, the licensee discovered that the turbocharger air inlet temperatures caused a derating
of the EDGs if outside ambient air temperatures exceeded a particular value.
Initially,- the licensee thought that the removal of insulation from the EDG exhaust manifolds, to support a modification, was the cause.
Subsequently, the licensee replaced the insulation and performed extensive testing to determine the heat profile within the room and the effects on the EDGs and related equipment.
Based on the testing performed by the licensee, it vas established that the additional heat load imposed by the removal of the EDG insulation was not the causative factor in the EDGs potentially operating outside their design basis. The elevated temperature problem was inherent in-the rooms since original construction.
The design of the EDG rooms does not facilitate j
radiant cooling.
Therefore, it cannot be concluded that the engineering evaluation made, as part of the original modification, was erroneous and contributed to the inoperability of the EDGs at elevated temperatures.
Licensee engineering action witn respect to this concern has been evaluated'
extensively by the NRC during 10S9 and 1990. Overall, the engineering investigation, evaluation, testing, and corrective actions have been strong. The results of related inspections are documented in NRC Inspection Reports 50-285/89-25, 50-285/89-32, 50-285/90-26, 50-235/90-30, and 50-285/90-34.
In a letter dated September 12, 1990, the licensee provided the NRC's Office of Nuclear Reactor Regulation (NRR) with an in-depth and comprehensive evaluation of the EDG postaccident electrical loading and the effect of elevated temperatures on EDG operability. The evaluation is
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currently under review by NRR, and any additional issues concerning EDG operability, if identified, will be resolved between the licensee and NRR.
4.
Review of Corrective Actions for Violations (92702)
a.
(Closed) Violation 285/9032-01:
Failure to perform a TS required surveillance test.
The licensee was cited for the failure to comply with TS 3.7(1)c.iii, which requires a test be conducted to ensure that emergency loading of the EDG does not exceed the 2000 hr-kW rating of the EDGs.
The licensee had design calculations demonstrating that power
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requirements for the equipment, which is automatically loaded during a design basis accident, were within the 2000 hr-kW rating of the EDGs, However, the design calculations were not proceduralized nor were they required to be analyzed every refueling outage, as required by TS 3.7(1)c.iii.
The licensee completed a comprehensive study of EDG loading and the effects of elevated ambient air temperatures on EDG capacity and
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performance.
In this study, the licensee concluded that the loading of the-EDGs will be below the 2000 kW-hr rating for nonextreme ambient air temperatures.
This analysis, FC-90-062 dated September 12, 1990,
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is currently under review by NRR.
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Additionally, on March 13, 1991, the licensee issued Surveillance Test SE-ST-DG-0001, " Verification of Diesel Generator Loads," to comply with the requirements of TS 3.7(1)c.iii. The inspector reviewed the procedure and found that it met the intent of the TS.
The implementation of the surveillance test, coupled with enhancements made to modification procedures, should ensure that the licensee maintains an accurate accounting of EDG loading, b.
(Closed) Violation 285/9038-01:
Safety injection relief valve
setpoint error.
In September 1990 it was noted that the relief velves protecting
. safety injection piping, bounded by the safety injection tank discharge isolation valves and the first check valve downstream of the iso'ation valves, did not conform to the design requirements of USAS B31.7, in that the relief setpoints of relief Valves51-278,
-279, -280, and -281 were found to be set at 395 psig, whereas the piping they serve was designed to only 250 psig.
As discussed in NRC Inspection Report 50-285/90-38, the licensee took prompt corrective actions. The corrective actions included performing a stress analysis of the ef fected piping to verify the suitability of the increased pressure, visually verifying that the effected piping did not exhibit any permanent deformation, and reviewing operational history to determine if -the piping could have been overpressurized.
The analysis indicated that the piping could be upgraded to a higher pressure limit.
The visual inspection _found no damage, and operational history records indicated it was highly unlikely that the piping ever experienced pressures in excess of its design rating. Accordingly, the licensee issued Safety Analysis for Operability (SAO) 90-10 to document operability of the effected piping.
The inspectors evaluated the SA0 in NRC Inspection Report 50-285/90-38 and found it to be satisfactory.
For long-term corrective actions, the licensee committed to hydrostatically test the piping during the 1992 refueling outage to achieve a 395 psig design rating. Additionally, the licensee committed to enhance its inservice inspection program by requiring the testing of all safety-related relief valves at a minimum frequency of once every 5 years.
The licensee incorporated these requirements into-Procedure PE-ST-VX-3001, "ASME Section XI Code Relief Valve Test
. Procedure." During verification and validation of this procedure, the relief valve setpoints were compared to existing design basis documentation to ensure compatibility.
By memorandum dated March 15, 1991, from the Manager, Station Engineering to the Manager, Nuclear Licensing and Industry Affairs, this verification and validation had been completed.
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5.
LER Followup (92700)
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The following event reports were reviewed to verify that reportability requirements were fulfilled, corrective actions were accomplished, and actions were taken to prevent recurrence, a.
(Closed) LER 89-017:
Raw water system outside its design basis.
This LER addressed a situation where the raw water system was operated outside its design basis when a check valve, necessary to prevent system backflow, failed.
The licensee's handling of this situation,
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short-term corrective actions, and inspector analysis of licensee actions is documented in NRC Inspection Report 50-285/89-28.
The licensee's engineering staff provided outstanding support for the operations staff in developing compensatory actions that allowed the system to be safely operated.
The root cause was determined to be lack of an inservice testing program for the raw water check valves.
The licensee previously committed to establish a check valve testing program as a result of NRC Information Notice SS-70, "Chect Valve Inservice Testing Program
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Deficiencies." This commitment was made via the licensee's Safety Enhancement Program (SEP), Item 43.
Licensee completion and NRC review of SEP Item 43 is documented in NRC Inspection Report 50-285/91-13.
b.
(Closed) LER 89-024:
Potential use of containment spray (CS) system outside design basis.
This LER identified the potential for use of the CS system outside of its design basis.
Engineering analysis revealed that the CS pumps and suction header were not originally designed for use as a backup to the low pressure. safety injection system (LPSI) for shutdown cooling, i
Such use was outside the pressure and temperature limits for the CS system. The TS, at the time, erroneously allowed the CS system to be used for shutdown cooling, as provided by FCS operating procedures.
Initial licensee corrective actions and inspector evaluation of the actions is documented in NRC Inspection Report 50-285/89-50.
In summary, it was determined that no immediate safety concern existed since-operational logs indicated that the CS system was never used in=
a configuration'that would have caused excessive thermal or~ pressure 1nduced stresses.
The 11mensee completed the following long-term corrective actions:
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Emergency operating procedures were revised to limit the conditions for use of a CS pump for shutdown cooling, in the event both LPSI pumps were inoperable.
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TS 2.3 was revised to delete interchangeability_among LPSI and CS
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pumps. TS 3.16 was amended to lower the test pressure on the-i
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shutdown cooling suction header.
- An emergency repair procedure, in the event of fire impairment of LPSI Pump SI-18, was incorporated into Abnormal Operating Procedure ADP-6, " Fire Emergency."
The licensee's response to Generic Letter 88-17, " Loss of Decay Heat Removal,"_-was supplemented to reflect appropriate use of the.
CS pumps for shutdown cooling and alternate means of decay heat removal _without the use of the CS pumps.
This was done via a letter dated June 8, 1990, i
The inspector _ verified that the long-term corrective actions had been completed.
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6.
Onsite Followup _of Events (93702)
a.
Loss of the Technical Support Center ITSC) Uninterruptible Power
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Sugply (UPS)
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On April 11,1991, at 10:40 a.m. _(CDT) the licensee declared a Notification of Unusual Event (NOVE) based on a significant loss of assessment capability for an accident.
. The licensee's electrical maintenance organization was conducting testing on the EDG power supply system that provides backup power to
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the TSC.
During this testing, all power supplies were lost to the
- TSC, Loss of the UPS system resulted in the complete loss of the plant's emergency response facility computer system and the associated safety. parameter display system. The apparent cause of the event was that a control switch was not placed in the test position, as' required by procedure, because of personnel error.
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In accordance with Emergency Plan Implementing Procedure.0SC-1,
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" Emergency Classification," Emergency Action Level 5.2, a NOUE was
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declared due to;the.significant loss of assessment capability.
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states of Nebraska and Iowa were notified, as well as the NRC Operations Center.
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The licensee restored _the UPS system at 10:49 a.m. and the NOUE was terminated at 11:05 a.m.
Incident Report (IR) 910096 was written to assess the event and recommend-corrective actions-The licensee's action in declaring the NOUE was proactive since Procedure OSC-1 required that a NOUE be declared if assessment information is not available from alternate sources.
However, the licensee decided to declare the NOUE and not delay the notification to determine if an alternate assessment capability was available.
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P b.
Main-Feedwater Regulating Valve Oscillation
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On April 22, 1991, a main feedwater regulating valve (FCV-1101) began to oscillate and caused swings in Steam Generator RC-2A level, Operators switched the valve control from automatic to manual mode, but found that they were not able to operate the valve.
Loss of control of the valve was caused by failure of an instrument air line that supplied air to the valve controller.
A feedwater regulating bypass valve (HCV-1105) was opened to the 35-percent position and the
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oscillations stabilized.
Valve FCV-1101 is equipped with a manual (handwheel) override, which operators took control of and opened the valve to the 44 percent position.
Steam generator level was maintained with the use of Valve HCV-1105.
The licensee decided to commence a plant shutdown to 35 percent to make any necessary repairs.
At 35 percent power, the level in-RC-2A was controlled by Valve HCV-1105, and FCV-1101 was isolated to facilitate repairs.
The licensee generated Maintenance Procedure pE-RR-VX-0408N, " Inspection and Repair of Non-Safety Related Fisher "EHD" Control Valves."
Following disassembly, it was determined that the probable cause of the failure was the installation of an incorrect plug (disc assembly)
during the 1988 refueling outage, The incorrect plug caused flow instabilities, which caused the valve to oscillate, and resulted in excessive wear to the carbon rings on the plug.
The valve oscillations became apparent when it was decided in early February 1991 to operate the plant _ for an extended period at
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l 75 percent power, The differential pressure and flow characteristics
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across the plug apparently were more conducive to vibration and oscillations at 75 percent than at 100 percent, since the oscillations were not experienced at full power.
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Tne licensee had eiperienced this problem in the past and, as corrective action, the valve vendor providtd a modified plug in 1983 with different characteristics, which resolved the vibration _and oscillation problems.
However, it appears that the vendor-continued to supply the old plug as the standard part for the valve and left it up to the customer to modify the plug, as required. As a result, a new plug was procured and installed in 1988 without being modified.
The licensee initiated-IR 910111 to describe-the incident-and required that a root.cause analysis-be performed to address the problem-of the incorrect component having been installed.
Since the licensee's root cause analysis program is handling this issue, this item will-be
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re,icoM by the inspectors when the licensee's review has been ctmpleted.
c.
Inadvertent Ventilation Isolation Actuation Signal (VIAS)
On April.30, 1991, while removing a label from 120-Vac Cabinet AI-40C, an electrician inadvertently tripped a normally closed breaker. This i
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breaker-feeds control room Panels AI-33A (process radiation monitors)
- and AI-45-(auxiliary coolant panel). The tripped breaker caused the loss of-power to Radiation Monitors RM-050 (containment air particulate) and -051 (containment gas) that are used to generate a containment radiation high signal (CRHS). A VIAS results directly from a CRHS.
The electrician immediately reset the breaker, but four containment isolation valves (PCV 742-E, -F, -G, and -H) closed upon the VIAS, as designed.
The operators reset the CRHS alarm and lockout relay, verified that all equipment operated, as designed, and reset all valves to their proper alignment.
The electrician was in the process of replacing-black background breaker labels with more easily read white background labels at the time of the event. When he attempted to scrape off the glue from one of the labels, his hand slipped-and the breaker was struck.
The inspector will review the licensee's corrective actions during-followup of LER 91-007.
Conclusions The plant operators promptly and ef fectively responded to a plant perturbation when a main feedwater regulating valve began oscillating.
Apparent personnel error resulted in a declaration of a NOUE and a VIAS.
7.
Operational Safety Verification -(71707)
a.
Routine Control Room Observations
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The inspectors observed operational activities throughout this inspection period to verify that-proper control room staffing and -
control room professionalism were maintained, and shif t turnover -
meetings were conducted in a manner that provided for proper communication of plant status from one shift to the ether.
Discussions with operators indicated-that-they-were aware of plant-status, equipment-status, and reasons for lit annunciators.
Control room indications of various valve and breaker lineups were verified
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for current plant conditions. The inspectors observed that all limiting conditions for operation were properly documented and tracked by control room personnel.
It was also noted-that the licensee-maintained the emergency notification system in an operational status, b.
-Plant Tours The inspectors routinely toured various areas of the plant to verify _
that proper housekeeping was being maintained. The inspectors noted-i -
an improvement in the housekeeping in the radiation controlled i
area (RCA).
Painting activities in the RCA were steadily progressing and the overall appearance was very good.
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On April 30, 1991, the inspector toured both EDG rooms to verify proper valve alignment for the fuel oil and starting-air systems.
No problems were noted during the tour.
c.
Radiological Protection Program Observations
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On May 9, 1991, during a tour of the RCA, the inspector noted an individual apparently working inside a contaminated area without the use of protective clothing.
There were two workers in Room 22 (east safety injection pump room) in the process of installing concrete a
anchors in E pillar for the addition of a pipe support. One of the
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workers was observed on a scaffold approximately 8 feet off the ground. The scaffold was built from the ground up and was located in a roped-off posted contaminated area..The worker was in street clothes. Another scaffold was observed outside of the posted area.
The inspect ( r questioned the second worker, who was outside of the
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posted area, about protective clothing requiremt nts.
The individual stated that i health physics (HP) technician had told them that as
long as thev ware above the contaminated area they did not need to wear proter.tive -lothing.
The individual stated that they used the second sctffold sa cross over to the scaffold that was located above the contaninated area. The roped-off area was around Cantainment Spray Pump SI-3C and included the side of the pillar that the individual was working on.
The inspector lef t Room 22 to locate an HP technician and question the use of street clothing in that area. Af ter a discussion with an HP technician and the shift HP supervisor, it was determined that, during the prejob briefing, the two workers had been instructed that-protective clothing was not requirea as long as they remained above the roped-off area.
However, the HP personnel decided that the controls present could be improved,
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On May 10, the inspector toured Room 22 again and spoke with an HP technician working in the room.
The HP technician had surveyed the area in question and was able to move the posted barriers to arourj the: containment spray pump, leaving the area under the pillar as a clean area.
In addition, signs were installed, at locations.around-the scaffold, prohibiting entry to the scaffold through the
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contaminated area.
L The_ licensee's actions on this matter was very prompt. Although the workers were observed obeying their radiation work permit and HP instructions, the licensee's actions helped to minimize the possibility of an individual becoming contaminated.
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Security Program Observations The-inspectors observed various aspects of the security program.
Personnel, packages, and vehicles were noted to be properly searched before entering the protected area.
It was noted that guards had been e
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posted when security doors were open for plaret ar.tivities.
Professionalism of the security guards was observed to be excellent,
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e.
Interpretation of 10 CFR 50.72 Reporting Requirements j
On April 15, 1991, NRR issued an interpretatier. of 10 CFR Part 50.72
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reporting requirements.
The interpretation sas pro.?pted by a letter sent to the NRC by the Virginia Electric'and Power Company (VEPCO).
'VEPCO stated in the letter, dated December 18, 1990, that they intended to no longer notify the NRC, pursuant to Part 50.72(b)(2)(vi),
solely because state and local government agencies were notified of an event occurring at onelof their facilities.
The definition of government agency, as used by VEPCO, was from Part 50.2, which does not include state and local governments.
The NRC's Office of General Counsel reviewed VEPCO's position and stated that the term " government agencies," as used in Part 50.72, was intended to include state and local, as well as federal agencies.
The inspector notified the FCS onsite licensing organization of the interpretation, in addition, the interpretation was presented at the monthly exit meeting on May 17,.1991, f.
Licensee Audit and Surveillance of Control Room Activities The inspector reviewed a selected quality assurance (QA) audit and surveillances of control room _ activities to determine if the scope and schedule were being met and whether findings were being resolved in a timely manner.
The following audit and surveillances were reviewed:
Audit Report 29, "Shif t Operations and Tagout_," dated January 12, 1990 QA surveillance Report 02-90-2, " Plant Operations - Operator Duties - Control Room,"--dated October 26, 1990
QA Surveillance _ Report-02-91-1, " Plant Operations --Operator-Duties - Control Room," dated February 13, 1991-QA Surveillance Report Al-89-2, " Control _ Room Drawiegs," dated
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QA Surveillance Report Al-90-1, " Control Room Drawings," dated July 6, 1990'
QA Surveillance-Report 04-90-2, " Temporary Modification Control,"
dated October 5, 1990 J
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QA Surveillance Report 04-91-1, " Temporary Modification Control,"
dated April 24, 1991 The inspector noted that the audit and surveillances were comprehensive in scope and the findings indicated good attention to detail.
In addition, Corrective Action Reports (CAR)90-415 and 91-047, generated from two of the surveillances, were reviewed.
The licensee's corrective actions resulting from the CARS were timely and addressed the QA department concerns.
Conclusions Control room activities were properly performed.
Radiation protection and security personnel performed their duties in a good manrer.
A licensee QA audit and surveillances, and resulting corrective actions, were found to be comprehensive and prompt.
8.
Maintenance Observations (62703)
On April 16, 1991, the inspector witnessed a portion of the replacement of Raw Water Pump AC-10C.
The replacement was performed under Maintenance Work Order 911254 using Procedure MM-RR-RW-001, " Removal and Installation of Raw Water Pumps."
The inspector observed that all required equipment was properly tagged out of service and that the procedure used was the latest revision.
In addition, it was observed that the procedure was properly followed.
The coordination between the maintenante personnel performing the job and the quality control (QC) inspector was good since very little time was lost due to waiting for the QC inspector to witness an activity at a hold point.
The inspector noted that good care was taken to prevent damage to any of the equipment that was being installed.
In addition, the plant personnel safety coordinator was present to help ensure employee safety.
Conclusions Maintenance was performed in accordance with procedures.
Good coordination was noted between maintenance and QC.
9.
Surveillance Observations (61726}
a.
Power Rance Safety Channel Bistable On April 17, 1991, a system engineer discovered that a TS-required surveillance had not been performed.
TS 3 1, Table 3-1, Item 1.c, requires that, for the power range safety channels, a surveillance of the internal test signal to verity trips, alarms, permissives, and auctioneered circuits be performed monthly.
The permissives are the Level 1 Bistable (15 percent power) for which no monthly surveillance procedures existed.
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The bistable was installed to disable the axial power distribution and the loss of load trips and enable the startup rate trip when power
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decreases below 15 percent.
The converse is true when power is increased above 15 percent.
Tl licensee declared the bistable inoperable but did not declare-the-trip units inoperable since these are routinely-tested monthly, which would indicate that the bistable had worked to enable these Units.
Thus, there was not a safety concern at the plant's current operating power level.
The licensee contacted Combustion-Engineering'and was informed that other plants do not perform this testing on a monthly basis..The licensee subsequently decided that a TS discrepancy existed and that a TS change or interpretation was necessary.
The inspectors will follow up on this issue during the review of LER 91-008.
In addition, until this issue is resolved, the inspectors will monitor licensee action when power is reduced below 15 percent.
b.
EDG ? Testing On Ma.' 15, 1991, the inspector witnessed the monthly testing of EDG 2 per Frecedure OP-ST-DG-0002, " Diesel Generator 2 Check." This check is performed monthly to satisfy TS requirements and provide a high degree of assurance of EDG. reliability.
Essentially, the ED's was started, brought up to full speed, fully loaded, and then run for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while essential parameters were monitored and recorded. The-inspecto' noted that the test was performed by a licensed operator who was dedicated to the testing effort. No anomalies were noted during the performance of Procedure'OP-ST-DG-0002.
In parallel with the performance of Procedure OP-ST-DG-0002, the licensee performed acceptance testing of Modification FC-90-073,
" Installation of Air Conditioning Unit for EDG Static Exciter Panels."
This modification was required since the-licensee discovered that heat buildup within the static exciter cabinet could. lead to failure of the electrical components in the cabinet.
In the summer of 1990, the
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voltage regulator within the cabinet failed, rendering EDG 2-inoperable. This event is discussed in NRC Inspection Report 50-285/90-32.
The event led to the initiation of a forced plant shutdown due to the potential for the common-mode failure of the
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EDGs.
In conjunction with the manufacturer, the licensee determined that static exciter' electrical components could fail at cabinet temperatures of 122 F.
Due to heat buildup from the operating EDGs, this limit corresponds to outside ambient air temperatures of 103 F and 100 F for EOGs 1 and 2, respectively.
The modification testing ensured that the thermostatically controlled air conditioning unit properly cycled and maintained the cabinet internals within an
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approximate 4 F band. Additionally, the test verified that _the control-room received an alarm if cabinet temperatures reached 115 F,
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which could occur in the event of failure of the air conditioning unit.
The testing was observed by the inspector.
It was found to be
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properly executed with the use of an approved procedure.
The test was executed _by system, electrical design, and test and performance engineering personnel.
No anomalies were noted with the procedure or its implementation.
For 2 years, the licensee has been evaluating the thermal characteristics of the EDG rooms and the effects of elevated temperat;ces on EDG performance.
This year, the licensee began using advanced thermographic analysis techniques.
The techniques should prove beneficial in evaluating vulnerabilities in EDG performance-during the upcoming summer months.
Overall, the licensee performed EDG testing in a controlled and thorough manner.
The ongoing investigative testing illustrated sound engineering and a commitment to EDG reliability and plant safety.
c.
Csntrol Room High-Efficiency Particulate Air Filter Testing On May 15, 1991, the inspector observed testing in progress per Procedure OP-ST-CR-0001, " Control Room Filtered Circuit Operation."
The purpose of this test is to satisfy TS requirements that ensure operability and reliability of the overall control room filtered ventilation system, an engineered safety feature.
This particular test requires operation of the system in the filtered-air makeap mode for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> each month and that the system be monitored for abnormalities. The-inspector independently verified that the system was properly aligned for the filtered-air makeup mode for Train A.
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No disc _repancies were noted.
Conclusions Surveillances were performed according to procedures with good attention to detail.
The licensee's discovery of a missed TS surveillance for a power range safety channel bistable indicated another benefit of the systems
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engineering program.
10.
Exit Interview The inspectors met with Mr. W. G. Gates (Division Manager, Nuclear
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Operations) and other members of the licensee staff on May 17, 1991.
The meeting attendees are listed in paragraph 1 of this inspection report.
At-this meeting, the inspectors summarized the scope of the inspection and the findings.
During the exit meeting, the licensee did not identify as proprietary, any information provided to, or reviewed by, the inspectors.
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ATTACHMENT
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l Acronyms and initials
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- corrective. action report
- CRHS
- containment-radiation high signal CS
- containment spray FCS-
- Fort Calhoun Station-EDG
- emergency diesel generator i
IR
- incident report HP
- health physics
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LER
- licensee event report
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- low pressure safety injection
- NOUE
- Notification of Unusual Event NRC
- Nuclear Regulatory Commission
- NRR
- Nuclear Reactor Regulation OPPD
- Omaha Public Power District psig pounds-per square inch gauge
- QA
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quality assurance QC quality. control-
- RCA radiation controlled area SA0
. safety analysis for operability SEP
- Safety Enhancement Program TS'
- Technical Specification TSC
' UPS
--uninterruptible power supply Vac
- alternating current voltage VEPC0 - Virginia Electric and. Power System-VIAS.
ventilation isolation actuation signal
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