IR 05000285/1994024
| ML16343A732 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/10/1995 |
| From: | Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML16343A288 | List: |
| References | |
| 50-285-94-24, NUDOCS 9502220313 | |
| Download: ML16343A732 (26) | |
Text
<<c U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-285/94-24 Operating License:
DPR-40 Licensee:
Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
P.O.
Box 399, Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska Facility Name:
Fort Calhoun Station Inspection At:
Blair, Nebraska Inspec8tion Conducted:
November 14 through December 27, 1994 Inspectors:
R. Hullikin, Senior Resident Inspector E. Collins, Team Leader L. Smith, Senior Resident Inspector, Arkansas Nuclear One G. Werner, Reactor Inspector M. Schlyamberg, NRC Contractor Approved:
>am
.
o nson,
>e
,
rogect rane ate Ins ectio Summar d:
ill i
i th i
t ng the licensee's identification that component cooling water (CCW) temperature could =Bexceed the maximum design value during accident conditions and that the control room air conditioning (A/C) units would likely be rendered inoperable.
Results:
~0enati nns
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The operators appropriately assessed the operability determination provided by engineering and initiated the Technical Specification required plant shutdown.
The Notification of Unusual Event was declared as required by the emergency plan emergency action level for the required plant shutdown (Section 2. 1. 1).
En ineerin
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The licensee identified a significant design basis deficiency which would result in both control room A/C units being rendered inoperable during a main steam line break (HSLB) inside the containment building or a large break loss of coolant accident (LOCA).
The failure would result
r
from a CCW temperature in excess of the A/C units'esign operating limit.
This design deficiency was identified as a result of their initiative to perform a Service Water System Operational Performance Inspection (Section 2).
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A violation was identified for the failure to assure that the control room air conditioning unit design modification correctly translated the design basis specifications for assuring system operability during certain design basis accidents.
As a result, the control room air conditioning units that were purchased and '.nstalled were not capable of operating within the component cooling water maximum temperatures following a postulated main steam line break inside the containment or a large break loss of coolant accident.
This condition has existed since 1988 (Section 2. 1,8).
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A second violation was identified for a Failure to implement the established procedures for the documentation and evaluation of the design deficiency.
Plant personnel responsible for addressing the control room A/C units'perability concern did not notify operations when the concern was first raised and did,not independently ensure that a prompt operability determination was performed.
This resulted in the plant operating at 100 percent power for an additional month outside its design basis (Section 3).
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A comprehensive safety analysis for operability was developed based on an excellent coordination of operations and engineering inputs to provide compensatory measures to ensure a control room A/C unit would remain operable following a OBA.
The analysis was well supported by their plant specific probablistic risk assessment (Section 2).
Hang ement Oversi ht
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Hanagement was proactive in implementing the Service Water Operational Performance Inspections Appropriate resources were dedicated to the'reparation and conduct of the activity (Section 3).
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Hanagement was not effective in assuring that the operability concerns were promptly reviewed and acted upon in accordance with the established corrective action process, once they were notified on October 17, 1994 (Section 3).
Summar of Ins ection Findin s:
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Violation 285/9424-01 was opened (Section 2. 1.8).
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Violation 285/9424-02 was opened (Section 3).
DETAILS
INTRODUCTION On November 14, 1994, the licensee notified the NRC that CCW temperatures could exceed the maximum limit described in the Updated Safety Analysis Report and the Design Basis Document.
This report contains the results of two inspection efforts which assessed immediate safety concerns and the review of the actions prior to the licensee's determination of the problem.
IMMEDIATE SAFETY CONCERNS On November 14, 1994, the inspectors assessed the immediate safety concerns and the licensee's actions to justify continued operation following the determination that CCW temperatures could exceed the maximum described limit following a DBA.
The licensee determined that the higher CCW temperatures would render the control room A/C units inoperable.
2.1
~lno erab1e Control Room A
C Units 2.1.)
Oescri tion of Initial Event On November 14, 1994, the licensee's engineering organization determined that the control room A/C units would be rendered inoperable during a large break LOCA or an HSLB inside containment.
The loss of the A/0 units would be caused by increased CCW temperature through the A/C units'ater cooled condenser and waterside economizer which would result in the rupture discs on the Freon system actuating, emptying the system of Freon.
The major increase in CCW temperature would result from the heat transfer from the containment cooling units following a DBA.
After A/C failure, the control room temperature would continue to rise until the offsite power low signal (OPLS) relays, lockouts, and sequencers would become inoperable at 105'F room temperature, which would correspond to an internal cabinet temperature of 120'F.
The OPLS is an engineered safety feature (ESF) which load sheds the safety-related 4160-volt busses, starts the emergency diesel generators, and energizes the busses following a loss of offsite power with a safety injection actuation signal present.
The licensee entered Technical Specification 2. 15(3), which required that the reactor be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the OPLS inoperable.
Technical Specification 2. 15(3)
was entered at 1:49 p.m.,
A Notification of Unusual Event was declared at 2:55 p.m.
as required by the licensee's emergency plan for a plant shutdown required by the Technical Specifications.
At 4:30 p.m.,
a reactor power decrease was initiated at a rate of 1 percent per hour.
Concurrently, the licensee was reviewing the design deficiency and working toward mitigating its significance.
Engineering and operations sub equently developed compensatory actions that could be taken to mitigate the significance of the design deficiency and would permit continued plant operations until the refueling outage.
The Plant Review Committee met to
consider the proposed alternatives to the shutdown.
The Plant Review Committee approved the compensatory measures based on an engineering analysis and Operations Hemorandum 94-07.
These actions provided for placing one of the two control room A/C units (VA-46A) in "STOP" and tagging closed the CCW valves to the unit.
This would preserve the one unit for control room cooling.
The memorandum required shutting down the operating control room A/C unit after a plant trip.
However, if no DBA had occurred and CCW temperature remained below 90'F and was not rising, then this unit could be restarted.
If a
DBA had occurred, then the following actions were to be taken by operators:
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Maximize raw water (RW) flow through available heat exchangers.
Have only two CCW pumps operating to reduce CCW flow rate and enhance heat transfer.
Reduce the containment cooler heat load on the CCW system when containment pressure is-reduced to an operator judged acceptable level.
Continue controlling CCW heat loads until CCW temperature is reduced to less than 90'F.
Then the protected control room A/C unit can be placed into service to maintain control room temperature.
In the event that CCW temperature cannot be controlled, operators would then enter the appropriate abnormal operating procedure (AOP) for either controlling temperature or establishing a raw water backup.
The appropriate AOP's would be AOP-ll, "Loss of Component Cooling Water,"
and AOP-13,
"Loss of Control Room Air Conditioning."
The licensee also imposed more restrictive equipment operability requirements than specified in the Technical Specifications for the control room ventilation, raw water pumps, and CCW heat exchangers.
The following actions were defined:
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The loss of the operating control room A/C unit (VA-468) would require entering a 24-hour Limiting Condition for Operation (LCO). If the unit could not be repaired within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the licensee would enter Technical Specification 2.15(3)
and be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The Technical Specification did not have an action statement on either A/C Unit VA-46.
However, Technical Specification 2. 12. 1 stated that, if the temperature in the control room cabinets exceeded 120'F, and could not be reduced below 120'F in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then the reactor must be put into hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
An inoperable CCW/RW heat exchanger would require entering a 14-day LCO.
The Technical Specification did not require any action for one inoperable heat exchanger.
An inoperable RW pump would require entering a 24-hour shutdown LCO.
The Technical Specification would require entering a 7-day shutdown LC The operations memorandum was supported by an operability evaluation and a
CFR 50.59 review.
The
CFR 50.59 review concluded that an unreviewed safety question existed without the actions defined in the operations memorandum.
At 7:28 p.m.,
the licensee stopped the power reduction at 97.5 percent.
The Notification of Unusual Event was terminated at 7:50 p.m.
2. 1.2 S stem Descri tion The CCW was a closed system of demineralized water which provided the cooling medium for many plant heat loads using three CCM pumps.
The major heat loads following a recirculation actuation signal were the shutdown cooling heat exchangers (AC-4A and -4B), containment air cooling units (VA-7C and -7D),
and containment air cooling and filtering units (VA-3A and -3B).
CCW was cooled by RW through four RW/CCW heat exchangers using four RM pumps.
During normal operation, one CCW pump, one RW pump, and one heat exchanger would be in service.
The control room was cooled by two redundant A/C units (VA-46A and -46B).
Normally only one unit would be in operation.
There were also two control room filtering units (VA-64A and -64B).
2.1.3 Descri tion of Concern The containment cooling units and the containment spray system provided the capability to reduce containment temperature and pressure after a large break LOCA or an HSLB inside containment.
The containment cooling units were only required during an HSLB.
A containment isolation actuation signal would have been received following either of these two accidents.
The containment cooling and filtering units would have automatically started and CCW would have supplied these two units.
In addition, both control room A/C units and fans would have automatically started, including the CCW supply to the A/C Unit VA-46 Freon condensers.
To evaluate the CCW temperature transient, the licensee performed Calculation FC-06304 and determined that within 3 minutes of a large break LOCA or HSLB, concurrent with a loss of instrument air, the CCW temperature would exceed 130'F for certain initial configurations of the RW and CCW system"..
The compressors for the control room A/C Units VA-46A and -46B would trip at 106 F and, at a
CCW temperature of 130'F, Freon pressure would actuate the system rupture disk, releasing the Freon into the control room complex.
The licensee determined that the Freon would not pose a habitability concern.
The loss of the Freon, however, would disable the A/C units.
Without cooling, the control room temperature would increase within 30 minutes to the Technical Specification limit of 105'F.
The licensee would enter AOP-13 to attempt to reduce the control room temperature.
Some of the methods outlined in the procedure would be to shut down unnecessary control room heat loads and open the control room doors, However, opening the control room doors would have potentially created a radiological hazard during a
LOC o
The 105'F control room temperature would not have created a severe
'habitability concern by itself for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but would have affected ESF equipment.
The licensee determined that the first to be affected would have been the OPLS.
The OPLS equipment would have been affected by an internal cabinet temperature of 120'F, which corresponds to a 105'F control room temperature.
If a loss of offsite power occurred at the beginning of the accident, OPLS would perform its functions.
However, if the loss of offsite power were to have occurred at 30 minutes after the loss of control room cooling, then the OPLS might not have performed its design function.
2. 1.4 En ineerin Calculations The following licensee engineering calculations were used in the resolution of this concern:
Calculation FC-06236,
"Post-DBA Heat Load on CCW System as a Function of CCW Return Temperatu'. e" The objective of this calculation was to determine the maximum post-DBA heat load on the CCW system as a function of CCW return temperature, which is the temperature exiting the CCW/RW heat exchangers.
This calculation showed how the maximum heat load on the CCW system changes if the CCW system heats up in a post-DBA situation.
Calculation FC-06304,
"RW and CCW Initial Temperature Rise for LOCA-Haximum Safeguards" The objective of this calculation was to estimate the RW and CCW temperatures during the early stages of a large break LOCA with full safeguards actuation.
Calculation FC-06308,
"CCW Return Temperature Maintainable at RAS with 50'F River Temperature" The objective of this calculation was to determine whether the RW system could support a
CCW temperature of 90'F or less during the post-RAS period of a LOCA with river temperature of 50'F or less.
The calculation considered one RW pump feeding four RW/CCW heat exchanger s, which simulates one inoperable RW pump and a failure of an emergency diesel generator.
In addition, the calculation assumed three RW pumps feeding three RW/CCW heat exchangers, which simulates the one RW pump inoperable and the failure of a RW isolation valve to open on one RW/CCW heat exchanger.
Calculation FC-06311,
"Control Room Heat Gain Without Air Conditioner VA-46A or -B Cooling"
The objective of this calculation was to determine the control room heatup rate between the present time and the spring 1995 refueling outage.
The following assumptions were made:
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Both A/C Unit VA-46 compressors are shut off at time 0.
One of the A/C Unit VA-46 fans is running at time 0 but is shut off at time 20 minutes and not restored.
The A/C Unit VA-64 charcoal filter has a
9-KW heating element that is energized the entire time.
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The control room temperature is 80'F, with 50 percent relative humidity at time 0.
The outside air temperature is 60'F maximum with 50 percent relative humidity.
2.1.5 Safet Anal sis for 0 erabilit On November 18, 1994, the licensee approved Safety Analysis for Operability (SAO) 94-02,
"Control Room Air Conditioners VA-46A/B, ESF Lock-outs, Relays, and Sequencers."
The SAO concluded, based upon engineering calculations, that the control room temperature would not exceed 105 F if outside ambient air temperature remained less than or equal to 60'F, with no control room A/C units in service.
All control room equipment would remain operable under this scenario.
Additional conditions were established in the event that outside air temperature were to exceed 60'F before the 1995 refueling outage, scheduled to begin on March 11.
The licensee used the results of probabilistic risk assessment (PRA) models to conclude that control room equipment would be operable for outside temperatures greater than 60'F when the duration of these temperatures was less than or equal to 250 cumulative hours.
This assumed that the operating VA-46 unit fan would be stopped at or before 20 minutes after a
DBA since the fans also provided a
heat load.
In addition, if the Missouri River temperature was less than or equal to 50'F the licensee concluded that the CCW temperature could be reduced sufficiently to allow starting of the protected VA-46 unit.
The licensee also evaluated whether elevated CCW temperatures would affect other safety-related equipment cooled by CCW.
The conclusion was that these pieces of equipment would be operable with river temperature less than 60'F, three RW pumps available, and three RW/CCW heat exchangers available.
The SAO provided operational restrictions during normal plant operation.
These restrictions were:
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no more than 250 total hours that outside air temperatures reach 60'F.
This would be logged and tracked in the control room;
g Ir
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one RW pump may be inoperable indefinitely.
One RW/CCM heat exchanger may be inoperable For 14 days.
However, the RW pump and heat exchanger could not be inoperable at the same time; two RW pumps may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; RW flow must be maintained through three RW/CCW heat exchangers; river temperature must be less than or equal to 50'F; one A/C Unit VA-46 must be protected; the A/C Unit VA-46 in service may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as long as control room temperature was below 80~F; and
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control room temperature must be below 80'F.
The SAO stated that, if any of the above conditions were not met, Technical Specification 2. 15(3) would be entered, which required a shutdown of the plant.
In addition, the SAO also provided for actions to take if a large break LOCA or an HSLB inside containment were to occur.
These actions involved stopping operating A/C Unit VA-46 at or before 20 minutes into the accident and, if outside air temperature exceeds 60 F, then restarting of protected A/C Unit VA-46 would be desired.
In order to bring this unit into service, the following actions were defined:
maximize both the number of RW pumps and RW/CCW heat exchangers in service; ensure a maximum of two operating CCW pumps in order to maximize heat transfer capability;
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when sufficient containment spray flow was verified, then stop all containment cooling and filtering units except for one filtering unit (VA-3A or -3B).
The remaining filtering unit could be stopped if CCW temperature did not decrease to less than 90'F;
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start protected A/C Unit VA-46 if CCW temperature was below 90'F and decreasing; and
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if the control room temperature reached 100'F and CCW temperature had not decreased to less than 90'F then the operators should consider entering either AOP-11 or -13.
On November 29, the licensee submitted a letter to the NRC Region IV Regional Administrator detailing the operational requirements that they imposed on themselve. 1.6 Re ortabilit Process NOD-(P-31, "Operability and Reportability Determinations,"
is the procedure that provides overall guidance for determining operability and reportability of nonroutine events and conditions.
Section 7.3.2 stated that an initial operability determination shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> even though complete information may not be available.
The licensee determined that both control room A/C units would be inoperable following a large LOCA or NSLB inside containment.
In addition, since the A/C units are support equipment for ESF
'omponents, the OPLS was also declared inoperable'owever, the operability determination stated that the plant could continue to'perate with the recommendations previously described.
The inspectors questioned the licensee as to when an SAO would be written and approved since this is the licensee's document for justifying continued operation.
guality Procedure NOD-gP-22,
"Preparation and Approval of a Safety Analysis for Operability (SAO)," states that the SAO is a comprehensive and documented evaluation of a safety or operability concern which established the bases for the continued safe operation of the plant.
The procedure did not require that the SAO be approved within a certain time frame, but a prior revision of this procedure had established a time requirement of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
2.1.7 Use of PRA The licensee used PRA models to determine the risk involved by taking interim actions prior to the 1995 refueling outage.
The models used were the following:
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Change in core damage probability versus numbers of hours that outside ambient air temperature is greater than 60'F.
This resulted in a change of approximately 1.00E-06 for up to 250 cumulative hours of outside temperature greater than 60'F.
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Change in core damage probability versus time that outside ambient temperature is greater than 60'F with standby A/C Unit VA-46 available.
This resulted in a change of approximately 5.00E-07 for up to 250 cumulative hours of outside temperature greater than 60 F.
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Same as above but with A/C Unit VA-46 unavailable.
This resulted in a change of approximately 1.00E-05 for up to 250 cumulative hours of outside temperature greater than 60'F.
2. 1.8 Technical Review The inspectors reviewed the licensee's calculations and evaluation of the effects of CCW temperature exceeding the design value of 120'F.
The inspectors found that the original CCW maximum design temperature apparently did not consider the consequences of the maximum heat rejection from containment during an accident and nonconservatively bounded the heat
, ~
-10-rejection from the CCW system through the RW/CCW heat exchangers.
Also, the licensee did not establish the basis for the maximum design CCW temperature (120'F) during the design basis reconstitution program.
This situation was further exacerbated by the use of 90'F as a maximum cooling (CCW) water temperature in the purchase specification for the replacement of the. original A/C units (Modification HR-FC-81-51),
which took place in 1988.
This was not consistent with the design parameters of the CCW system.
During the service water self-assessment, the licensee identified weaknesses related to the single failure analysis which had previously been performed on the CCW system.
The inspectors concluded that the lack of a thorough and rigorous single failure review factored in not identifying the maximum safeguards condition earlier.
The inspectors questioned the impact of this condition on the CCW pump net positive suction head (NPSH).
The NPSH was of particular concern, since it is inversely related to CCW temperature.
The inspectors'oncern about the NPSH was compounded by the uncertainty of the ability to maintain nitrogen overpressure in the CCW surge tank.
The nitrogen supply was not single failure proof and some of the pressure retaining components connected to the tank were not safety related.
The licensee responded that the NPSH would be adequate within the RW and air temperature limits specified in LER 94-010,
"Potential Accident Scenario Involving Loss of Control Room Air Conditioners."
The licensee indicated that a more comprehensive evaluation of the NPSH, including loss of the overpressure, as well as the other potential impacts on the CCW system, would be addressed during the detailed CCW transient analysis.
The inspectors concluded that the most limiting impact of high CCW temperatures was the control room A/C failure and that there was no immediate safety concern with the continued operation of the plant.
The inspectors noted that, neither the Updated Safety Analysis Report nor,Design Bases Document had considered the impact of the maximum safeguards equipment temperatures on the CCW system.
Criterion III, Appendix B of 10 CFR Part 50 requires that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.
In addition, Criterion IV, requires that measures shall be established to assure that applicable regulatory requirements, the design basis, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services.
The failure to appropriately translate the design basis for the CCW system into the control room A/C unit design modification specifications is a violation (298/9424-01).
As a result, the control room A/C units that were purchased and installed in 1988 were not capable of operating within the CCW system design basis temperatures following a postulated main steam line break inside the containment or a large break loss of coolant acciden ~ t I
-11-
PROBLEM IDENTIFICATION AND EVALUATION The inspectors reviewed the sequence of events leading to the licensee's identification of the maximum safeguards equipment temperature concerns and the associated documentation and evaluation.
The inspectors found that, in the spring of 1994, the licensee assembled a
"preinspection" team in anticipation of an NRC service water operational performance inspection team.
In about April, it was decided that Fort Calhoun Station would perform a self-assessment in lieu of a team inspection and the NRC's inspection was postponed.
As part of the preinspection team, the licensee questioned the maximum CCW design temperature of 120'F.
No calculation could be found to support the number.
In March 1994, the licensee had requested that the architect engineer evaluate the consequences of CCW temperature exceeding 120'F.
The results of this evaluation were documented in Stone and Webster Calculation 16472.8030-PH-l.
Revision 1 of this calculation, dated October 3 (available onsite about October 5) included information that the control room A/C units would trip at a cooling water temperature of approximately 102'F.
The licensee continued to pursue the problem, but the incident report operability/reportability processes were not entered at this time.
On October 17, senior licensee management was briefed of a potential problem with CCW temperature.
On October 26, the reportability procedure was entered with the drafting of a memorandum to plant operations.
No incident report was generated.
Licensee engineers apparently focused on the postaccident CCW transient temperature and could not find any analysis indicating what the temperature would do.
The licensee used iterative hand calculations and discovered that the, temperature would rise very quickly and that the control room A/C unit Freon rupture discs would likely actuate.
On November 14, the control room A/C units were declared inoperable.
Stone and Webster Calculation 16472.8030-PM-l, Revision 1,
was onsite October 5 and indicated that the A/C units would trip at a temperature of approximately 102'F.
This information indicated a design deficiency since the maximum design CCW temperature was 120'F.
This design deficiency was not entered into the corrective action system and no prompt operability determination was performed.
Inspectors reviewed Standing Order SO-R-4 and found that Section 2.4.5 stated that violations of established system design bases (Updated Safety Analysis Report or Design Bases Document) or externally imposed regulations are types of incidents requiring an incident report.
The inspectors also reviewed Nuclear Operations Division Procedure NOD-gP-31, Revision 8, and found that Section 6.7 assigned the responsibility For prompt operability and reportability determinations of events or conditions to the Sh'.~t Supervisor.
Section 6. 16 stated that all
"NOD, PED, and NSD" personnel are responsible for identification to appropriate personnel of nont outine events and conditions, including nonconformances, which may require
J
-12-operability and reportability determinations, by initiating an incident report per Standing Order R-4.
Section 7. 1 stated that determinations of operability and reportability will be performed using the guidance in this procedure and will be implemented through Standing Order R-4, "Station Incident Reports,"
and Standing Order R-11, "Notification of Significant Events."
However, certain situations may meet prompt
CFR 50,72 reporting requirements or other regulations; for this reason, the shift supervisor must be informed of these events concurrent with or ahead of the incident report generation.
Cognizant personnel shall provide information and recommendations necessary for the shift supervisor to determine prompt reportability.
The inspectors concluded that the design deficiency indicated by Calculation 16472.8030-PH-I, Revision 1,
was a condition that violated established system design basis and warranted an Incident Report on October 5, under the requirements of Standing Order R-4.
Also, Procedure NOD-gP-31 directed the initiation of an incident report for conditions which may require an operability determination.
As a result of not initiating the incident report, the shift supervisor was not notified of the condition and a prompt operability determination was not performed for conditions where it was known or extremely likely that the control room A/C units would trip under accident conditions.
In addition, potential problems were informally evaluated.
An incident report was written on November 15 after the systems were declared inoperable.
This is a violation (285/9424-02).
f>
ATTACHtIENT
PERSONS CONTACTED 1. 1 Licensee Personnel
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Andrews, Division Hanager, Nuclear Services Blome, Supervisor, Corporate Quality Assurance Boughter, Acting Manager, Station Engineering Cavanaugh, Licensing Engineer Chase, Manager, Fort Calhoun Station Cook, Supervisor, Station Licensing Eurich, Supervisor, Design Engineering Gasper, Manager, Training Gates, Vice President, Nuclear Jaworski, Manager, Station Engineering Kusek, Manager, Nuclear Safety Review Group Lakin, Nuclear Safety Review Group Specialist Orr, Manager, Quality Assurance and Quality Control Patterson, Division Manager, Nuclear Operations Phelps, Acting Manager, Production Engineering Sandhoefner, Shift Supervisor Sefick, Manager, Security Services Skiles, Acting Manager, Design Engineering Tesar.
Manager, Corrective Actions Tesarek, Supervisor, Simulator Services Tills, Assistant Plant Manager, Operations Trausch, Manager, Nuclear Licensing and Industry Affairs Van Sant, Mechanical Engineer, Design Engineering
Denotes personnel that attended the exit meeting on December 13, 1994.
Denotes personnel that attended the exit meeting on December 16, 1994.
+
Denotes personnel that attended the exit meeting on December 27, 1994.
In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
EXIT HEETING An exit meeting was conducted on December 27, 1994.
During this meeting, the inspector reviewed the scope and findings of the report.
The licensee acknowledged the inspection findings.
The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.