IR 05000285/1996014
| ML20135C659 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/03/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20135C600 | List: |
| References | |
| 50-285-96-14, NUDOCS 9612090030 | |
| Download: ML20135C659 (18) | |
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l ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
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Docket No:
50-285
l License No: DPR-40 l
l Report No:
50-285/96-14
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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 l
Facility:
Fort Calhoun Station l
l Location:
Blair, Nebraska
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t Dates:
October 6 through November 16,1996
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Inspectors:
W. Walker, Senior Resident inspector V. Gaddy, Resident inspector K. Johnston, Senior Resident inspector i
F. Brush, Resident inspector S. Campbell, Project Engineer l
l Approved:
W. D. Johnson, Chief, Project Branch B Attachment: SupplementalInformation
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9612090030 961203
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PDR ADOCK 05000285
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EXECUTIVE SUMMARY j
Fort Calhoun Station NRC Inspection Report 50-285/96-14
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i This routine announced inspection included aspects of licensee operations, engineering, maintenance and plant support. The report covers a 6-week period of resident inspection.
Operations Operators drained the reactor coolant system to midloop in a controlled and
effective manner. The inspectors noted good communications and control by the licensed senior operator during the draindown (Section 01.2).
The operators responded effectively to the unexpected power loss to the toxic gas
monitors (Section 01.3).
The inspectors noted that poor communications and coordination between a
licensed senior operator and an auxiliary operator resulted in misalignment of the fuel transfer carriage (Section 04.1).
The inspectors concluded that licensee corrective actions were appropriate
following a failure of operations personnel to follow procedures and properly bypass the steam generator low pressure trip (Section 08.2).
Maintenance The maintenance activities and surveillance tests observed were conducted in a
controlled and professional manner (Section M1.1 and M1.2).
A minor weakness was noted in foreign material exclusion controls during
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maintenance on a component cooling water heat exchanger (Section M4.1).
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l Engineering department control and knowledge of the containment ventilation
design basis were good. The physical condition of the filter train and cooler inspected was good (Section E2.1).
Engineering personnel appropriately evaluated a change in the alert setpoint for a j
containment stack radiation monitor (Section E7.1).
j The inspectors concluded that licensee corrective actions were approp iate
following a failure of licensee personnel to ensure the accuracy of the engineering i
analysis and safety analysis used to change the fuel handling equipment procedure (Section E8.1).
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Plant Support The inspectors observed that additional efforts to maintain good housekeeping
inside containment were needed (Section R1.1).
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l Report Details i
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Summary of Plant Status
i The Fort Calhoun Station was in its 16th refueling outage during this inspection period. A
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j major activity during the outage was eddy current testing of both steam generators. Based i
on test results,36 tubes in Steam Generator RC-2A were plugged and 21 tubes in Steam i
Generator RC-2B were plugged. This brings the total number of plugged tubes to j
92 and 76 for Steam Generators RC-2A and RC-2B, respectively.
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Another may activity was reconstitution of failed fuel. The licensee determined that there i
was a total of /8 leaking fuel pins in 28 fuel bundles. Nine of the assemblies with failed
pins were reconstituted with stainless stee; pins and placed back into the core, r
1. Operations
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Conduct of Operations j
01.1 General Comments (71707)
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The inspectors frequently observed ongoing plant operations. In general, the conduct
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of operations was professional and safety-conscious. The inspectors noted effective
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Specific events and noteworthy observations are detailed in the sections below, i
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01.2 Reactor Coolant System Drainina to Midloop i
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Insoection Scope (71707)
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The inspectors observed the licensee drain the reactor coolant system inventory to j
midloop for maintenance.
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Observations and Findinge
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On October 11,1996, the inspectors observed licensed operators drain the reactor
coolant system inventory to midloop to uncouple the reactor coolant pumps, install j
steam generator nozzle dams and perform other maintenance. The inspectors noted l
that the draindown was conducted using Attachment 3 of Ol-RC-2A, " Reactor j
Coolant System Fill and Drain Operations." Due to the significance of this evolution, j
operators were given a pre-evolution briefing by management. Management briefed j
operators concerning general background information while draining to mid-loop, i
including the risk involved and problems encountered by industry during this j.
evolution. Management discussed crew responsibilities and reenforced expectations
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for the draindown. Also discussed were potential problems that could occur, criteria j
for terminating the draindown and contingency plans.
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Prior to draining to midloop, inspectors verified that the licensee had completed all procedural prerequisites and discussed procedural precautions with the licensee.
j The inspectors noted that the licensed senior operator maintained command and i
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control during the draiodown. Good communications were exhibited by all personnel.
The inspectors noted that the draindown was successfully completed in accordance
with the procedure.
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Conclusions
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The management briefing of control room operators prior to draining to midloop was
good. The brief covered personnel responsibilities, reenforced expectations, and discussed potential problems that could occur. The licensed senior operator maintained good command and control while draining to midloop. Good communication was demonstrated by operators.
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01.3 Loss of Power to Toxic Gas Monitors YlT-6286B and -6287B a.
Inspection Scooe (71707)
The inspectors reviewed the circumstances surrounding an inadvertent power loss to Toxic Gas Monitors YlT-6286B and -62878.
b.
Observations and Findinas On November 3,1996, inverter 2 blew a fuse and a power loss to Instrument Bus Al-42B for the toxic gas monitors occurred while changing the inverter power source from the bypass transformer to the normal source,125 V DC Bus 2. The operators entered Abnormal Operating Procedure 16, Section Vil, " Loss of Instrument Bus Al-42B," as a result of the event and wrote Condition Report 199601381 to document the unexpected power loss. Power was restored to the toxic gas monitors when the power source was returned to the backup transformer.
The licensee had previously replaced the degraded bypass transformer with a transformer from a different manufacturer and was in the process of completing testing on the new bypass transformer when the power loss occurred. The inspectors concluded, after reviewing Engineering Change Notice 96-406 which evaluated the "non like-for-like" replacement of the transformer, that the power loss was not attributed to incompatibility with the new transformer and the inverter. The ECN concluded that the form, fit and function between the old and the new transformer were identicalif not better characteristics for the new transformer.
Additionally, the inspectors reviewed the abnormal operating procedure and confirmed that the licensee met the procedure requirements in response to the inss of power to the instrument bus.
Following troubleshooting of the inverter circuitry, the licensee replaced two electronic cards (an oscillator and a logic board) and a silicon controlled rectifier.
Inverter 2 was returned to service on November 7 after successful testing.
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Conclusions The inspectors concluded that the licensee's response to the unexpected power loss j
was appropriate. The inverter repairs were successfully completed and the invertir was returned to service.
Operator Knowledge and Performance 04.1 Fuel Transfer Carriaae Damaae a.
Inspection Scone (71707)
l The inspectors reviewed the licensee's actions associated with a loss of ventilation in the auxiliary building and containment, and the subsequent closure of the fuel transfer canal isolation valve which resulted in misalignment of the fuel transfer carriage.
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Observations and Findinas On October 15,1996, while installing the cover on an overcurrent relay, an electrician bumped and actuated the relay. Tripping of the relay caused a loss of all 480 Vac buses powered from Diesel Generator 2. The loss of 480 V power resulted j
in a loss of ventilation in both the containment and auxiliary building, and a loss of -
l lighting in the spent fuel pool area. At the time, the fuel transfer carriage was located in the fuel transfer canal. With the loss of the ventilation systems, a
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manometer effect could have occurred from the difference in pressure between the j
auxiliary building and the containment and the fuel transfer canal could have
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l overflowed.
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An auxiliary operator was directed by personnel in the control room to close Valve
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FH-11, the transfer canal isolation valve. Attempting to close the valve with the -
transfer carriage in the transfer canal caused the carriage to become misaligned in the canal. This prevented the carriage from being moved and Valve FH-11 could not
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be completely closed. Even though FH-11 was not completely closed, the transfer i
canal did not overflow and the transfer carriage was realigned in the transfer canal.
l The licensee deteimined that the following contributed to the event.
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The operator did not verify the position of the carriage due to the perceived
urgency to close the valve.
There were no interlocks, alarms or signs to warn of a potential interface
problem between the carriage and valve.
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Valve FH-11 is normally hard to operate. When the valve came into contact
j with the carriage, the operator did not recognize the problem.
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The control room personnel did not know the status of the carriage when the i
operator was told to close Valve FH-11.
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The fuel transfer canal did not overflow even though Valve VH-11 was not closed,
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and the fuel transfer carriage was realigned with the operating track.
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Conclusions.
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The inspectors concluded that poor operator communications and coordination led to.
the fuel carriage misalignment.
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Miscellaneous Operations issues (92700)
08.1 (Closed) LER 50-285/96-008: ventilation isolation actuation signal due to high containment activity. On October 11,1996, with the plant in Mode 5 for refueling, l
mechanics entered the containment to remove the steam generator primary side
manways. The containment purge system was in service when the ventilation isolation actuation signal occurred. The containment purge isolated as designed to
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i limit the release of radioactivity from the containment to less than the limits specified
in 10 CFR Part 20 and the Technical Specifications.
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l The licensee identified the cause of the event to be inadequate procedural controls I
during an activity that breached the primary system boundary. The ventilation actuation isolation signal occurred due to an increase in containment airborne j
activity. The increased activity was caused by the opening of the cold leg manways j
with a high efficiency particulate air filter ventilator in service on the hot leg j
manways. The result was a rapid dispersal of xenon gas into the containment
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atmosphere. A contributing factor was higher than normal fission product activity j.
due to fuel element failures.
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l The licensee was reviewing and updating procedures to ensure appropriate controls are in place to preclude a ventilation isolation activation signal from occurring dunng
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j maintenance and other operational evaluations inside containment.
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The inspectors reviewed the licensee's recommended corrective actions which t
appeared to be appropriate.
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08.2 (Closed) LER 50-285/96-007: unplanned reactor protection system actuation while cooling down. On October 5,1996, with the plant in Mode 3, operators were j
cooling down the reactor coolant system to begin the refueling outage. All control
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. rods were inserted. Operators were cooling down using Operating Procedure OP-3A,
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" Plant Cooldown," Attachment 3, " Reactor Coolant System Cooldown from Mode 3 l
to Mode 4/5."
j At approximately 1:15 p.m., when pressure dropped below 1700 psia, operators j
blocked the pressurizer pressure low signal as required by Step 7 of the attachment.
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5-Steps 8 and 9, respectively, directed operators to bypass the steam generator low pressure signal when the reactor protective system pretrip actuated and to block the engineered safety features steam generator low signal when the pressure in both generators was less than 550 psia.
At 1:23 p.m., the steam generator low pressure signal and reactor protective system pre-trip annunciators actuated which indicated that steam generator pressure was low enougn to bypass and block these signals. Operators acknowledged the alarms using a common acknowledgement button. At 1:32 p.m., with steam generator pressure below 550 psia, operators blocked the engineered safety features steam generator (ow signal (Step 9).
At 1:36 p.m., one of the steam generator low pressure trip units actuated, operators then realized that the step to bypass the reactor protective system steam generator low pressure trip had not been performed (Step 8). As operators were acquiring the keys to bypass the steam generator low pressure signal, a second steam generator low pressure trip unit actuated. This resulted in a reactor protective system trip signal. The control rods were already inserted into the reactor. Since the diesel generator test switches had previously been opened in accordance with Procedure OP-3A, the diesels did not start. However, the initiation of the rear. tor protective system trip signal did cycle the matrix relays for equipment that normally would have been required to actuate. After the initiation of the trip signal, operators installed the steam generator low pressure bypass keys and reset the reactor trip signal. Failing to bypass the steam generator low pressure trip as required by the procedure is a violation of 10 CFR 50, Appendix B, Criterion V. This licensee-identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.I of the NRC Enforcement Policy (285/96014-01).
The licensee performed a root cause analysis of the actuation and determined that the licensed senior operator failed to complete Step 8 of the procedure. Corrective actions included a revision to Procedure OP-3A to minimize the potential of a similar
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event. Specifically, the procedure was revised to require a senior reactor operator to conduct a brief and assign a licensed operator the responsibility of bypassing the steam generator low pressure trip and blocking the steam generator low pressure
signal. The procedure was also to be revised to ensure Step 8 and 9 were kept on the same page. These actions were to be completed by November 21,1996. The licensee also indicated that the procedure would be enhanced by using place keepers j
to assist operators in ensuring that critical steps were completed. The inspectors concluded these actions were appropriat. _ -. -
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i 11. Maintenance j
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M1 Conduct of Maintenance
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M1.1 General _ Comments
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Inspection Scone (62707)
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The inspectors observed portions of the following work activities:
Diesel Generator Refueling Outage Inspection
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MOVATS Testing of Valve HCV-1185
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Observations and Findinas The inspectors found the work performed under these activities to be professional
and thorough. All work observed was performed with the work package present and
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in active use. The inspectors frequently observed system engineers monitoring
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i progress and quality personnel were present whenever required by the procedures.
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Conclusions b
The maintenance activities observed were conducted in a controlled and professional j'
manner, i
M1.2 Surveillance Activities
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a.
Insoection Scope (61726)
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The inspectors observed portions of the following surveillance activity:
J Channel Calibration of Containment High Pressure Switch B/PC-742-2
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Observations and Findinas
The' inspectors noted that this surveillance test was performed in accordance with
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the procedure. The surveillance procedure was present and in use during the l
observation. Communications between personnel performing the test were good.
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Conclusions The surveillance activity observed by the inspectors was completed in a controlled manner and in accordance with the procedur. _ =.
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M4.1 Removal of Tube in Component Coolina Water Heat Exchanaer l
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Inspection Scope (62707)
The inspectors performed routine walkdowns of the auxiliary building and assessed
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l work in progress. In particular, the inspectors observed preparations for the removal l
of a tube in a component cooling water heat exchanger.
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Observations and Findinas l
l The inspectors approached the job site for the t.omponent cooling water heat l
exchanger tube extraction several hours after w 3rkers had removed both heat l
exchanger end bells. The inspectors found that the raw water inlet to the heat exchanger which comes in at the bottom of the water box, was not covered. The l
inspectois questioned the workers present if it was a practice to install foreign material exclusion covers for these conditions. The workers responded by indicating it was a good idea and began to f abricate a cover. Subsequently, another worker i
returned to the job site and retrieved a pre-fabricated foreign material exclusion cover i
I from a box of pre-staged tools.
The inspectors reviewed the maintenance work order and Procedure PE-RR-CCW-0100," Disassembly, Cleaning and Repair of Component Cooling Water Heat Exchanger - Raw Water Side," and noted that there were no steps which addressed the installation of a foreign material exclusion cover.
The inspectors reviewed the procedure for foreign material exclusion (SO-M-10) and noted that it did not require a foreign material exclusion cover if a postwork inspection was performed. Procedure PE-RR-CCW-0100 did require a postwork closeout inspection, however, the procedure did not specify an inspection of the raw i
water inlet pipe. The inspectors discussed this issue with mechanical maintenance personnel and the system engineer. The inspectors were informed that a closeout inspection was intended to be performed on the raw water inlet pipe. The inspectors j
fourd this to be acceptable, i
in response to the inspectors' concerns, the licensee planned to review this procedure and add appropriate steps to ensure adequate foreign material exclusion
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controls. The inspectors considered 11., se proposed actions to be appropriate.
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Conclusions The inspectors noted minor weaknesses in the control of foreign material exclusion during maintenance on a component cooling water heat exchanger.
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l 111. Enaineerina E2 Engineering Support of Facilities and Equipment i
E2.1 Review of System Drawinas and Desian Basis for Containment Coolers a.
Insoection Scone (37551)
l The inspectors performed a walkdown of the containment air recirculation filter and l
cooler trains with the system engineer to assess system condition. Additionally the inspectors reviewed system drawings and design basis documents. For a selection of components, the inspectors reviewed documentation supporting the component's ability to perform its design function under design basis accident conditions.
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Observations and Findinas l
l The inspectors found that the licensee's design basis documents, with a few i
l exceptions, addressed the ability of the components to perform design basis
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functions under design basis accident conditions. The inspectors observed that the l
design basis document did not have verification that the pre-filter mist eliminators l
and a fiberglass mesh pad could withstand high temperatures. The system engineer subsequently contacted the vendor and determined that the components were l
designed for the high temperature environment. Additionally, the design basis I
document did not address the design requirements for the damper actuators which were required to change position under accident conditions. The system engineer was able to identify documentation which supported the ability of the dampers to perform design functions. Engineering personnel planned to revise the system design basis document to reflect this documentation.
The inspectors noted that pre-operational testing of the fan and motor for this system had verified that the motor would not be substantially impacted if it were running and atmosphere was abruptly changed from ambient to beyond design conditions.
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l However, the associated test record indicated that the motor had tripped when it j
was started with design basis conditions. Since these conditions appeared to be a more severe load on the motor the inspectors questioned whether the design basis load conditions on the motor, had been adequately considered by the licensee.
Engineering personnel subsequently reviewed the test data and concluded that the motor trips had occurred as a result of the power arrangement at the test facility.
l Additionally, system performance curves were reviewed and compared to motor data and breaker protection and the licensee determined that, in the installed condition,
there was reasonable assurance that the motor could start and operate under design basis accident conditions.
During the field walkdown, the inspectors observed that the filter train and cooler fc4 Fan VF3A were in good condition.
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Conclusions The containment ventilation design basis documentation, with a few minor exceptions, referenced documentation which established that the fan cooler components could perform intended functions under design basis conditions. The physical condition of the filter train and cooler for Fan VF3A was good.
E2.2 Review of Updated Safety Analysis Reoort Commitments l
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A recent discovery of a licensee operating their facility in a manner contrary to the Updated Safety Analysis Report description highlights the need for a special focused review that compares plant practices, procedures, and/or parameters to the Updated Safety Analysis Report descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the Updated Safety Analysis Report that related to the areas inspected. The inspectors verified that the Updated Safety Analysis Report wording was consistent with the observed plant practices, procedures, and/or parameters.
E7 Coality Assurance in Engineering Activities l
E7.1 Review of incident Reoort 940224: Wrgna Alert Setooint Value for Containment Stack Monitor RM-051 a.
insoection Scoce (37551)
The inspectors selected one of 30 incident reports that were over 2 years old and reviewed the licensee's evaluation of the incident report.
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Observations and Findinas incident Report 940224 documented, on June 13,1994, that Step 7.5.2 of Calibration Procedure CH-ST-RM-0053," Auxiliary Building Exhaust Stack / Containment Gas Radiation Monitor, RM-051, Primary Calibration," calibrated Containment Stack Monitor RM-051 at an alert concentration setpoint higher than the calculated concentration of Xe-133 released after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from a 1 gallon per minute reactor coolant system leak. Originally, the monitor was used as a
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contbinment stack monitor but its function was later changed to detect a reactor l
coolant system leak before break in the containment building. With the alert setpoint (
set above the calculated concentration, the monitor would be unable to perform this i
function.
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The inspectors reviewed Calibration Procedure CH-ST-RM-5100, " Containment Gas i
Radiation Monitor RM 051 Primary Calibration," and found that the required setpoint
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l was significantly lower than the alert setpoint listed in the 1994 incident report. A
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change in the Updated Safety Analysis Report, which lowered the value for average j
fission and corrosion product activity in the reactor coolant with 1 percent f ailed fuel,
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l However, the inspectors identified that a correction f actor included in Calculation
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FC-05462 was not included is the calibration procedure. The licensee stated that this factor applied to effluent nonitors as a result of a corrective action item in response to Licensee Event Report 77-17," inadequate Effluent Monitor Calibration l
Procedures," and not to leak before break monitors. The factor was therefore l
omitted from the calculation. After reviewing the licensee event report, the l
inspectors concluded that, as mentioned above, this monitor no longer functioned as an effluent monitor and therefore the omission of the factor was acceptable.
The licensee stated that a low priority was assigned to this incident report while the
licensee reduced a backlog of approximately 50 incident reports. The licensee stated that the backlog has been reduced and anticipated the incident report would be l
closed after the refueling outage.
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Conclusion The licensee appropriatel/ evaluated a change in the alert setpoint in Procedure CH-ST-RM-053. Further, the delay in closing the incident report, due to
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previous incident report backlogs, was acceptable.
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E8 Miscellaneous Engineering issues (92700)
E8.1 (Closed) LER 50-285/96003: fuel movement without control room ventilation in l
operation. On October 8,1996, the licensee determined that during the 1995 refueling outage, fuel had been moved without the control room ventilation filtration system in operation.
I in May 1990, the licensee issue.i control room dose calculations. These calculations assumed that the control room ventilation filtration system would be in operation before any fuel movement in containment, before any irradiated fuel movement or new fuel movement over irradiated fuel in the spent fuel pool area. In coincidence
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l with the completion of the control room dose calculations, the licensee initiated a j
procedure change to add a requirement to Procedure OI-FH-1, " Fuel Handling i
Equipment Operation," for a control room filtration unit to be operating in the filtered
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makeup mode whenever fuel was being muved. On July 19,1990, this procedure change was approved with a note that engineering would perform an analysis to see if this requirement could be removed, in February 1991, design engineering stated that the requirement could not be relaxed without an additional analysis. In October 1993, design engineering issued Memorandum PED-FC-93-2923. The memorandum stated that the control room
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ventilation filtration system was no longer required to be operating while moving fuel, in May 1994, Procedure Ol-FH-1 was again changed to remove the requirement of having the control room ventilation filtration system operating while moving fuel.
I In April 1996, while performing research to support a license change related to the
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control room heating ventilation and air conditioning system, the licensee noted that the engineering analysis used to change Procedure Ol-FH-1 in May 1994, did not provide an adequate basis for the change. On April 25,1996, the licensee concluded that the change was not adequately supported by an analysis and that during the 1995 refueling outage they had moved fuel without the control room ventilation filtration system in operation. The licensee then notified the NRC pursuant to 10 CFR 50.72(b)(2)(iii)(D) that fuel had been moved without the control room ventilation filtration system in operation.
i in May 1996, design engineering performed additional calculations which showed that the dose to the control room operators would be below the maximum control room dose assuming automatic initiation of the control room filtration system following an accident. Based on this new calculation, the notification made on April 25,1996, was withdrawn.
In October 1996, the licensee discovered that the analysis used to support withdrawing the notification credited Radiation Monitors RM-52 (Ventilation l
Discharge Duct Monitor) and RM-62 (Ventilation Duct Gas Monitor) as being l
redundant. However, the licensee determined that these radiation monitors were l
powered by the sarre power supply and subject to silgte failure. If the power supply l
failed, both radiaticn monitors would f ail and the control room filtration system would not automatically indate. Since the procedural change was not suppwted by an accurate analysis, the licensee resubmitted the withdrawn nc.tificatior, on October 8, 1996.
The licensee performed an investigation and concluded that the engineering analysis and the safety evaluation used to support removing the requirement of having the l
control room ventilation system in filtration was prepared without bdequate evaluation or review to ensure its accuracy. The licensee also concluded that the analysis that provided the basis for withdrawing the notification made on April 25, 1996, was inaccurate. Failing to ensure the accuracy of the engineerlag analysis and the safety evaluation used to change Procedure Ol-FH-1 is a violation of 10 CFR 50, Appendix B, Criterion Ill. This licensee-identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.I of the NRC Enforcement Policy (285/96014-02).
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Part of the licensee's corrective action included issuing an operations memorandum and changing Procedure OI-FH-1 to require that the control room ventilation filtration system be in operation prior to raoving fuel. On November 15,1996, the inspectors reviewed this procedure and noted that a step had been added in the precautions section to ensure the control room ventilation filtration system was operating prior to
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i fuel movement. However, a step in the prerequisites section only required the
control room ventilation system to be operable. The licensee indicated that the j
prerequisite would be changed to ensure that the control room ventilation filtration
system was operating prior to fuel movement. Procedure NOD-QP-3, "10 CFR 50.59 Safety Evaluations," had previously been revised to provide additional guidance on i
j the level of expertise and the use of documentation in the preparation of safety j
reviews. Additionally, the engineering analyses that formed the basis for the d
procedure change and withdrawing the initial notification were to be reviewed to i
identify and verify inputs and assumptions used and to verify that the results and i
operating configurations were appropriately incorporated into the Updated Safety
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Analysis Report and operating instructions using the 10 CFR 50.59 process and i
design basis documents. The inspectors concluded these actions were appropriate.
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IV. Plant SuDDort
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I R1 Radiological Protection and Chemistry Controls R1.1 Radioloaical Controls
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a.
Insoection Scope (71750)
On October 30.1996, the inspectors performed a tour of the auxiliary building and containment to verify that the material condition of safety and radiological components was adequately maintained.
b.
Ob::ervations and Findinos
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During the tour, the inspectors noted several areas inside containment which had an
accumulation of trash and other debris from ongoing maintenance activities. The
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inspectors discussed this with the licensee, and the nonessential material was i
removed from the containment.
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c.
Conclusion implementation of radiological controls as observed by the inspectors was generally characterized by careful radiological practices. The inspectors noted that additional efforts were needed to maintain good housekeeping inside containment around maintenance work areas.
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l-13-V. Manaaement Meetina l
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X1 Exit Meeting Summary The inspectors presented the inspection results to a member of licensee management at the exit meeting on November 18,1996. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
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i ENCLOSURE i
SUPPLEMENTAL INFORMATION
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i PARTIAL LIST OF PERSONS CONTACTED
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Licensee
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G. Bisnop, Assistant Plant Manager C. Brunnert, Manager, Quality Assurance and Quality Control J. Chase, Manager, Fort Calhoun Station R. Connor, Manager, Training S. Gambhir, Division Manager, Production Engineering I
J. Gasper, Manager, Nuclear Projects W. Gates, Vice President, Nuclear T. Jamieson, Supervisor, Radiological Operations R. Jaworski, Manager, Design Engineering, Nuclear B. Kindred, Supervisor, Nuclear Security Operations T. Matthews, Acting Manager, Nuclear Licensing E. Matzke, Station Licensing Engineer T. Patterson, Division Manager, Nuclear Operations R. Phelps, Manager, Station Engineering R. Short, Manager, Operations C. Simmons, Acting Manager, Nuclear Safety Review Group D. Spires, Manager, Chemistry M. Tesar, Manager, Corrective Action Group NRC V. Gaddy, Resident inspector W. Walker, Senior Resident Inspector I
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INSPECTION PROCEDURES USED
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IP 37551: Onsite Engineering IP 61726: Surveillance Observations l
!P E2707: Maintenance Observations
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IP 71707: Plant Operations
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IP 71750: Plant Support Activities IP 92700: Onsite LER Review ITEMS OPENED AND CLOSED Opened and Closed 50-285/96014-01 NCV failing to bypass the steam generator low pressure trip as required by procedure (Section 08.2)
50-285/96014-02 NCV failing to ensure the accuracy of the engineering analysis (Section E8.1)
Closed 50-285/96003 LER fuel movement without control room ventilation in operation (Section E8.1)
50-285/96007 LER unplanned reactor protection system actuation while cooling down (Section 08.2)
50-285/96008 LER ventilation isolation actuation signal due to high containment activity (Section 08.1)
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