IR 05000285/1996018

Insp Rept 50-285/96-18 on 961229-970208.Vioaltions Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support

ML20135E017
Person / Time
Site:	Fort Calhoun
Issue date:	02/28/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20135E014	List: IR 05000285/1996018 ML20135E011
References
50-285-96-18, NUDOCS 9703060203
Download: ML20135E017 (23)

Text

._ . . . . ..__ .. ._- _. _._. - - _ . _ _ . . _ . _ . - _ _ . .___..4._ ._ _ _ _ . . _ .

, .

i  :

!

'

.

'

l

!

ENCLOSURE i

'

l U.S. NUCLEAR REGULATORY COMMISSION

' i l

REGION IV  ;

i i

!  !

Docket No: 50-285  !

l License No: DPR-40 .

i i i Report No: 50-285/96018 i

Licensee: Omaha Public Power District l l Fort Calhoun Station FC-2-4 Ad l l P.O. Box 399, Hwy. 75 - North of Fort Calhoun  ! Fort Calhoun, Nebraska l- Facility: Fort Calhoun Station j i

l Location: Blair, Nebraska ]

i i

Dates: December 29,1996, through February 8,1997 .

l Inspectors: W. Walker, Senior Resident inspector l

! V. Gaddy, Resident inspector

]

Approved: W. D. Johnson, Chief, Project Branch B )

i-l l Attachment: Supplemental Information  ;

l l

'

<

i i

t.

t t

.

l l

l 9703060203 970220 PDR l

O ADOCK 05000285 l

! PDR j , ~ . . - - -- . - . - . - - - - - . . .--.- -. - .- _ - . - . ____ -

' .m

!

. EXECUTIVE SUMMARY  !

.  :

Fort Calhoun Station i t NRC Inspection Report 50-285/96-18

I I

This routine announced inspection included aspects 'of licensee operations, engineering, i maintenance, and plant support. The report covers a 6-week period of resident inspectio l 1 Operations -

t i

'

Operators responded effectively to the steam leak in the turbine building and !

'

s promptly declared an unusual event due to the personnel hazard associated with the

leak and the potential for the leak to worsen (Section 01.2).

j

• The inspectors noted a weakness in communication between operations personnel

! in that, following a surveillance test, control room ventilation was not returned to its normal configuration, as directed by the licensed senior operator (Section 01.3)

Maintenance

The maintenance activities and surveillances observed were conducted in a  !

controlled and professional manner (Sections M1.1 and M1.7).  ;

• The licensee identified three instances in which maintenance personnel failed to i properly follow configuration control procedures. This issue is unresolved pending further NRC review (Sections M1.2, M1.3, and M1.4).

• Documentation for postmaintenance testing of Valves HCV-921 and HCV-922 was missing. This issue is unresolved pending further NRC review (Section M1.2).

• The flinger ring on the outboard thrust bearing of a component cooling water pump was missing. Although this did not render the pump inoperable, this condition should have been documented by the maintenance technician on the maintenance I work order and elevated to station management attention for proper resolutio This issue is unresolved pending further NRC review (Section M1.5).

l

'

• Operators demonstrated a good questioning attitude in identifying a discrepancy during performance of a quarterly surveillance and initiated an appropriate procedure l'

change (Section M1.6).

Enaineerina

• The inspectors identified a potential discrepancy involving 480 volt cables inside a l cable tray in the east switchgear room. Additional followup will be performed  ;

(Section E1.1).  !

i Plant Sucoort  ;

l

• The licensee's radiological controls were generally well implemented (Section R1.2).

I l

,

.

.

-2-

The licensee identified that sampling of an effluent release via the auxiliary building ventilation pathway was degraded due to equipment necessary to successfully s;mple the etfluent being relocated without the knowledge of chemistry personnel

, (Section R3.1).

.

d W

1 i

l

]

.__

-

!

!

.

I Report Details Summary of Plant Status The Fort Calhoun Station began this inspection period operating at essentially 100 percent power and operated at that level throughout the inspection period with the following exception: power was reduced to 7 percent on December 31,1996, to repair a steam leak which could not be repaired without shutting the main turbine stop valve l. Operations 01 Conduct of Operations 01.1 General Comments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety conscious; specific events and noteworthy observations are detailed in the sections below. In particular, the inspectors observed excellent performance by the operating crews during the power reduction to repair a steam leak in the turbine buildin .2 Steam Leak in Turbine Buildina Inspection Scope (93702)

At approximately 11 a.m. on December 31, with the reactor at 100 percent power, l the licensee identified a steam leak in the basement of the turbine building. The '

inspectors observed the licensee's response to the event, including the declaration of notification of unusual even Observations and Findinas The leak was at a flange for a 1/4-inch orifice in a 1-inch bypass line off the common drain header for the main turbine control valve above seat drains. The leak presented a personnel hazard and could not be isolated without shutting the main turbine stop valves. The licensee determined that a rapid controlled reduction in power was necessary so that the turbine could be tripped cnd repairs made on the leaking orifice. The inspectors observed the control room activities during the power reductio The control room operators dispatched an auxiliary operator to check on the severity of the leak and to determine whether other equipment in the area was impacted by the steam leak, in addition, a plant-wide announcement was made for all personnel to stay clear of the turbine buildin ,

.

2-At 11.:28 a.m., the licensee declared a notification of unusual event based on the personnel hazard associated with the leak and the potential for the leak to worse ;

At 1:30 p.m., with power at 10 percent, operators tripped the main turbine, isolating the leak. The licensee terminated the notification of unusual event at 1:40 The inspectors observed that the licensee's overall response was quick and effectiv The control room staff exhibited strong command and control throughout the even Communications observed between auxiliary operr. tors stationed locally near the leak j and the control room were clear and included feedback to ensure that parties l understood the communications. After the steam leak was repaired, the licensee i initiated an investigation to determine the root caus I During the review of the potential cause, the licensee identified that the gasket material used (Garlock 3400, paper gasket) to seal the flange may not have been the correct application. The gasket was replaced with flexatallic, which is the material of choice for applications which are subjected to main steam pressure The licensee performed a root cause analysis and determined that craft personnel tightened the flanges with excessive force, causing over crushing of the gasket material. Also. the type of gasket material used contributed to the faibr The licensee was in the process of or has completed the following corrective actions: l l

• Evaluate the training and qualification of the worker who installed the gasket material on the flange which leake * Review the need to specify material for gasketed joints on drawings of i nonsafety-related component !

• Review need to specify material and torquing requirements on work instructions for skill of the craft activitie * Review all noninsolable pipinq that could cause a plant power loss or trip and determine need for installing isolation valve c. Conclusions The licensee's immediate response to the turbine building steam leak was excellent and minimized the effects of the steam leak, while providing timely notification of the unusual event. The inspectors reviewed the licensee's root cause analysis and proposed corrective actions and found them appropriat ,

.

.

-3-01.3 Control Room Walkdown Insoection Scope i71707)

The inspectors performed routine walkdowns of the control room.

t

! Observations and Findinas l

l On January 7,1997, the inspectors observed that an alarm, " control room A/C l system VIAS locked in," was activated on both trains of the Control Room l Ventilation Systems VA-46A and VA-468. This would indicate that the control room l ventilation system was in the filtered mode. The inspectors questioned the licensed l senior operator concerning why the alarms were in. The licensed senior operator stated that Surveillance Test IC-ST-RM-5001," Quarterly Function Test of Process Radiation Monitors," Revision 4, had been performed earlier in the day and that the control room ventilation should no longer be in the filtered mode and the alarms

! should have been reset. The licensed senior operator stated that an operator had I

been assigned to secure the control room ventilation from the filtered mode and to reset the alarms; however, the licensed senior operator had not verified that the VIAS L

alarm was reset and the filtered mode of control room ventilation secured. When the inspectors questioned the licensed senior operator about the alarms, the licensed senior operator appeared surprised that the VIAS alarms were stillin. Following the discussion with the inspectors, the licensed senior operator directed an operator to reset the VIAS signal at the local panel for the control room ventilation units and to secure the filtered mode of operation for control room ventilatio The inspectors discussed the above observations with the operations superviso The operations supervisor stated that the surveillance procedure would be revised to include a specific step for realigning of the control room ventilation units following performance of the tes Conclusions The inspectors concluded that operations personnel exhibited a weakness in communications in that, following a surveillance test, control room ventilation was not promptly returned to its normal configuration, as directed by the licensed senior l

'

operator. Also, the surveillance test contained a weakness in that no specific step directed operations personnel to realign the control room ventilation system to a normal configuration following the tes Operational Status of Facilities and Equipment 02.1 Enaineered Safetv Feature System Walkdown (71707)

The inspectors used inspection Procedure 71707 to walkdown accessible portions of I the high pressure safety injection system.

l l'

!

l l

!

.

.

l

, l-4- i i

l i

The inspectors roted that the material condition of the equipment was good. All supporti and seicmic restraints were properly anchored and in good conditio Also, all accessible valves were verified to be in the correct position, as required by procedure, and proper breaker position was observed at local electrical panel Miscellaneous Operations issues (92700) l l

08.1 (Closed) Licensee Event Report (LER) 50-285/95-005: plant trip due to operator error during diverse scram system testing. On August 24,1995, a licensed operator performing the diverse scram system actuation relay operability test, mistakenly ,

repositioned an incorrect switch from test to normal, causing a plant trip. This I violation was the subject of a Notice of Violation issued with NRC Inspection Report

.

50-298/95-1 The licensee determined the root cause of the event was inadequate attention to !

detail by the operator and that the error could have been prevented with additional self-checkin The licensee promptly implemented several actions following the plant trip. These l

. corrective actions included instituting the use of peer verification for operating !

i equipment from the control room that could cause a significant plant transien I I

The control room operators are now required to verbalize label information as an additional self checking technique prior to operating equipment. Also, Operations Policies and Directives OPD-4-15, " Conduct of OPEP Effectiveness / Operations Self-assessment Policy," was revised to include peer verification. The inspectors i directly observed several surveillances following this event and verified that the operators were performing control board manipulations according to the new polic .2 (Closed) LER 50-285/95-006: inoperability of Diesel Generator 1 discovered following a reactor trip. On August 24,1995, a reactor trip occurred due to operator error during the performance of the diverse scram system surveillance test. The

<

reactor trip provided a start signal to the diesel generators. Diesel Generator 2 started to idle speed (500 rpm) as designed, and Diesel Generator 1 started to full speed (900 rpm). This was not as designed. Issues related to this event were cited in a Notice of Violation issued with NRC Inspection Report 50-285/95-1 An operator error was identified in that, during the previous surveillance on Diesel

'

Generator 1, the generator voltage regulator control switch instead of the governor control switch was manipulated. There was no positive indication of governor positio The licensee promptly reviewed all diesel generator procedures to requ:re operators to use independent verification when the diesel governor is run back after a diesel generator is shut down. Also, a modification was completed that provided positive

.. - - . -- - - - _.-

.

.

.S_

,

'

governor position indication to operators, and procedure steps to provide for independent verification that the diesel governor had been run back were implemented. The inspectors verified by direct observation that the corrective

'

actions were implemented, i

08.3 (Closed) Unresolved item 50-285/96016-02: reportability determinations. This item remained opened pending the corrective action group providing the inspectors i additional information regarding their investigation into untimely reportability determinations. Standing Order SO'-R-2, " Corrective Action Program," required that reportability determinations be presented to the plant review committee for a final

'

reportability determination. The licensee determined that 18 reportability ,

determinations had not been presented to the plant review committee for a final l reportability determination. Although f ailing to have the plant review committee determine if an item was reportable is a violation of 10 CFR Part 50, Appendix B, Criterion V, the corrective action group stated that they were aware that this requirement was not being performed and were in the process of gathering data to

] determine the significance of the problem. Corrective actions included having the plant review committee review the 18 reportability determinations and revising l Procedure PED-WP-19 to ensure review of reportability determinations by the plant ;

>

review committee. This licensee identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.I of the NRC Enforcement Policy

,

(50-285/96018-01).

~

08.4 (Closed) ~v iolation 50-285/95024-01: failure to follow meintenance procedures when performing testing on the raw water / component cooling water heat exchanger l Maintenance Procedure PE-RR-CCW-0100," Cleaning and Flushing of Coolers l Supplied by the CCW System," directs operators to open the inlet component cooling water valve for the heat exchanger being tested. Operations personnel opened the inlet and outlet component cooling water valve, allowing component cooling water to be diverted away from the other heat exchangers in service. Operations crews were i trained on this event and a root cause analysis was performed by the Nuclear Safety Review Group, which was covered in training for the maintenance departmen Additionally Maintenance Procedure PE-RR-CCW-0100was revised. The inspectors reviewed the licensee's corrective actions and found them to be appropriat .5 (Closed) Violation 50-285/95004-01: f ailure to follow operations procedure when performing fuel movements. Operating Procedure OP-11, " Reactor Core Refueling,"

Revision 19, directed operators to perform core alterations as specified in Appendix A, " Fuel Movement Sequence." During the March 1995 refueling outage, the fuel handling machine operator removed the fuel element in Core Location C13 instead of the element at Core Location E15. Procedure OP-11 was revised requiring a second person at the fuel handling machine to verify, using the above water index system, the core location of the fuel handling machine prior to grappling an assembly. Additionally, operators were instructed on the importance of proper communications in providing bridge and trolley coordinates. The inspectors reviewed the licensee's corrective actions and found them to be appropriat .

)

'

l-6- l l

lI. Mainte_nan !

l M1 Conduct of Maintenance i

M1.1 General Comments 1 inspection Scooe (62707)

]

The inspectors observed portions of the following activities:

• Reinstallation of motor for Raw Water Pump AC-10A

• Lubrication and repacking of couplings for Main Feedwater Pump FW-4B

Flow control switch calibration for control room ventilation filtration unit (VA-648) Observations and Findinns The work observed by the inspectors was performed in a thorough and professional l manner. All work observed was performed with the work package present and in l

, active use. Maintenance craft were knowledgeable of the work being performe '

The inspectors verified that the test equipment used by the craft was calibrate !

l Conclusions Maintenance activities were generally completed thoroughly and professionally by knowledgeable craft personne M1.2 Unauthorized Modification to the Main Steam Line Radiation Monitor l

'

isolation Valves Insoection Scope (62707)

l The inspectors investigated the circumstances surrounding an unauthorized '

modification to the main steam line radiation monitor iso.ation valve Observations anc!Findinas On January 7,1997, design engineering identified that, in October 1996, maintenance personnel had replaced portions of Isolation Valves HCV-921 and HCV-922 for the main steam line Radiation Monitor RM-064 prior to the issuance of Engineering Change Notice 96-365 that would have authorized the replacemen Specifically, maintenance replaced the springs and a spiral pin on each valve. The . . _ _ _ _ __

.

.

I main steam radiation monitor was an offline monitor that measured the activity of steam by sampling either main steam header. The parts were replaced using 1 Maintenance Work Documents 962605 and 96260 !

l The engineering change notice was being prepared because springs in this type valve !

had failed at other sites and were identified as 10 CFR Part 21 failmos. The manufacturer recommended that the springs and spiral pins be replaced with a j different material. The engineering change notice was initiated on September 30, 1996, and still remained open, l

On October 22,1996, maintenance replaced the springs and spiral pins on the valves 1 4 with parts of a different material prior to the completion of the engineering change notice. On January 10,1997, the licensee completed an operability evaluation that concluded that, since the new material had superior resistance to corrosion and had high strength, ductility, and good mechanical properties, the valves were operabl Replacing and installing the springs prior to the completion of the engineering change notice may be a violation of Step 4.1.5(B) of PED-QP-2, " Configuration Change Control." This step required that configuration changes be authorized by one of several methods, one of which was an engineering change notice. This issue is j unresolved pending further NRC review (50-285/96018-02).

The operability evaluation also indicated that the vendor had performed tests that were not documented on the maintenance work documents. The inspectors j questioned if these tests were within the scope of the maintenance work documen '

The maintenance manager indicated that the vendor had performed the tests on the valves to ensure the valves were functional prior to leaving the site. The maintenance manager also indicated that the planner was aware of the work and the planner had initiated the proper documentation to perform the test. However, the documentation initiated by the planner that authorized the additional tests was missing from the work package. Without the documentation, the licensee and the inspectors could not definitively determine whether the tests performed by the

'

vendor were within the scope of the maintenance work document. This issue is also unresolved pending further NRC review (50-285/96018-03).

c. Conclusions The licensee identified that maintenance personnel had replaced portions of the main steam line radiation monitor isolation valves prior to the issuance of the engineering change notice that would have authorized the replacement. The licensee did not have the documentation to support the additional tests performed by the vendor on the isolation valve ~

l

i 8-

M1.3 Unauthorized Modification of a Shutdown Coolina Heat Exchanaer Outlet Valve Insoution Scoce (62707 and 35571)

The inspectors reviewed the circumstances involving an unauthorized modification to the Shutdown Cooling Heat Exchanger AC-4A component cooling water outlet valv Observations and Findinas On December 5,1996, the system engineer for the shutdown cooling system i identified that maintenance had performed an unauthorized modification to the component cooling water outlet valve for Shutdown Cooling Heat Exchanger AC-4 Specifically, maintenance had replaced the actuator cylinder of Component Cooling Water Outlet Valve HCV-484 prior to the completion of the engineering change notice that would have authorized its replacemen Earlier testing had confirmed that the stroke time for Valve HCV-484 had increased

{

and a decision was made to replace the actuator cylinder. On November 9,1996, i the system engineer initiated Engineering Change Notice 96-409 to evaluate the replacement actuator cylinder because the part number of the original actuator cylinder had been superseded and the new part was heavier and had a greater wall thickness. Maintenance Work Document 963534 was issued on November 20, 1996, and on November 22, maintenance obtained the part from the warehouse without the knowledge of engineering and installed the actuator cylinder prior to the completion of the engineering change notic In response to the unauthorized modification, the licensee performed an evaluation and concluded that, although the part dimensions were different, the valve was operable and could perform its safety functio The maintenance manager stated that the maintenance work document was to replace the actuator cylinder with a like-for-like replacement. However, since the part number had changed and the dimensions were different, maintenance planning should have flagged the maintenance work document to indicate that an engineering change notice was required prior to installing the valve. Installing the actuator cylinder into the component cooling water outlet valve prior to the completion of the engineering change notice may be a violation of Step 4.1.5(B) of PED-OP-2,

" Configuration Change Control," which established four controlled methods for changing the configuration of the plant, one of which was an engineering change notice. This is a second example of an unresolved item involving configuration control (50-285/96018-02).

_. ._ .- _ . . ._ . _ _ . _ _ . .._. _ _ _ _ _ . _ _ . . - _ . _ . _

.  !

.

. Conclusions i

Maintenance personnel installed an actuator cylinder on the component cooling water !

!

outlet valve prior to the completion of the engineering change notice that evaluated I the acceptability of the replacement part, j M1.4 Unauthorized Modification to the Safety Iniection & Refuelina Water Tank Vent Gasket ,

'! Insoection Scoce (62707 and 35571) l The inspectors reviewed the circumstances surrounding a modification to the safety injection and refueling water tank vent gaske I Observations and Findinas l

On January 2,1997, the licensee identified that maintenance had replaced the safety injection & refueling water tank vent gasket with a different size gasket than had !

been evaluated and authorized by Engineering Change Notice 95 387. The original gasket was made of asbestos and was no longer the preferred choice for gasket material. The engineering change notice specified that the replacement gasket be 1/16-inch thick, the same thickness as the original gasket. The engineering change notice indicated red rubber would be a suitable replacement material, i l

Using Maintenance Work Document 964427, maintenance began the process of replacing the gasket. Step 7.3 of the maintenance work document instructed maintenance to fabricate a new gasket from shop stock using red rubber stock material. The maintenance craft, in consultation with maintenance planning and engineering, decided to use a thicker 1/8-inch gasket to compensate for gaps in the flange. During the close out inspection, quality control questioned whether the thickness of the replacement gasket was appropriat The inspectors discussed this issue with engineering. The system engineer stated that the maintenance craft had originally replaced the gasket with a 1/16-inch red rubber gasket. Following the replacement, the vent flange leaked. The system engineer also stated that he, in consultation with the maintenance planner, provided direction to the maintenance craft to use the 1/8-inch gasket. The planner indicated he believed the thickness change was allowed by the engineering change notic Since the maintenance craft, at the direction of the system engineer and the maintenance planner, installed a gasket with a thickness different from that specified in the approved engineering change notice, the licensee performed an evaluation and concluded that the 1/8-inch gasket was acceptable. Failing to install the gasket that was specified by the approved engineering change notice may be a violation of Step 4.1.5(B) of Procedure PED-QP-2, " Configuration Change Control." The step required that configuration changes be authorized by one of several methods, one of i

,

-

- -_. - , _

- ._ _- . _ - - - . - . . - - .. ~. -- _

. -

_

_ - - . .. . . - ~ _

-

i e

- 10-i which was an engineering change notice. This is a third example of an unresolved item involving configuration control (50-285/96016-02). Conclusions

Maintenance personnel replaced the safety injection and refueling water tank vent

gasket with a thicker gasket than had been evaluated by an engineering change notic M1.5 Missino Flinaer on Component Coolina Water Pumo AC-3B Inspection Scoce (62707) '

While performing an inspection of the material condition of the component cooling water pumps, the inspectors observed a discrepancy on Component Cooling Water Pump AC-3 Observations and Findinas On January 22,1997, while performing an inspection of the component cooling water pumps, the inspectors noted that the flinger on the outboard (thrust) bearing .!

was missing. The flinger was de signed to prolong thrust bearing life by protecting the seal ring from water or debria if the outboard mechanical seal failed. The inspectors informed the system engineer of the discrepanc ,

l The system engineer investigated and determined that the missing flinger did not affect pump operability and initiated Maintenance Work Order 9700253 to install the ,

missing part, l in response to the inspectors' observations, the licensee determined that the rotating l assembly of the pump was replaced in October 1996. During the replacement, there was only one flinger on site. This flinger was installed on the inboard bearing. The outboard flinger was not installed, i Both the mechanic and the maintenance planner knew that the outboard flinger I would not be installed. Neither individualinformed their supervision or the system engineer that the flinger would not be installed. The maintenance manager indicated i

'

that the mechanic did not install the flinger to expedite completion of the work and to clear the limiting condition for operation. The inspectors reviewed the maintenance instructions that were used to perform the maintenance. Step 6.16.4 of Procedure MM-RR-CCW-0001," Inspection and Overhaul of Component Cooling Water Pump," required that the flinger be installed. This item is unresolved pending further NRC review (50-285/96018-04).

_ ,_

.

.

.

-11-l

!

l Conclusions I The inspectors identified that the flinger on the outboard (thrust) bearing was missing. The licensee determined that the flinger was not installed during the replacement of the pump's rotating assembly in October 199 M1.6 Surveillance Activities Insnection Scone (61726)

The inspectors observed portions of the following activities:

Procedure OP-ST-VA-3001, Ventilating Air System Quarterly Category A Valve Exercise Test

Procedure IC-ST-RPS-0026, Quarterly Functional Test of Thermal Margin / Low Pressure Channel A l l Observations and Findinas The inspectors noted that these surveillance tetts were performed in accordance !

with the procedures. The surveillance procedure was present and in use during the l observations. Communications between personnel performing the tests were goo J While observing the performance of Procedure OP-ST-VA-3001,the inspectors noted that the licensed operator performing the surveillance questioned the guidance !

provided in Steps 4 and 5 of Attachment 1 of the procedure. The note fcr Step 4 l required that two radiation monitors per channel be operable during the surveillance to actuate a containment radiation high signal. Radiation Monitors RM-052 and RM-062 were operable and capable of generating the containment radiation high signal. Since two monitors were operable, Technical Specification 2.15 did not have to be entere '

However, Step 5 of the surveillance directed the licensed operator to enter the Technical Specification for the radiation monitor. The licensed operator informed his supervisor of this discrepancy, and a procedure change was made to no longer require entry into the Technical Specification if two radiation monitors were available to actuate a containment radiation signal. Following the procedure change, the surveillance was successfully complete Conclusions Licensed operators exhibited a good questioning attitude by pointing out a discrepancy in a surveillance procedure. The surveillances observed by the inspectors were completed in a controlled manner and in accordance with the procedur _ _

-

l

.

-12-l l M8 Miscellaneous Maintenance issues (92902)

l l M8.1 (Closed) Followuo item 50-285/96001-03: failure of incore detectors. The i licensee, in the Updated Safety Analysis Report, committed to having 21 incore l detector strings operable. During Cycle 16, following the failure of five detector strings, the licensee completed a 50.59 evaluation to allow continued operation with as few as eight detectors, two per core quadrant. The inspectors verified that, during the Cycle 16 refueling outage, allincore detector strings were replaced. Also, the inspectors reviewed the root cause determination for the failed detectors. No new incore detector failures have been observed during the current operating cycle (Cycle 17).

Ill. Enaineerina E1 Conduct of Engineering E1.1 Potential Cable Routina Deficiencies Insoection Scone (37551)

During a routine inspection of the east switchgear room, the inspectors noted a discrepancy with the cables inside a cable tra ,

j Q1;servations and Findinos l l

On January 13,1997, while touring the east switchgear room, the inspectors noted that, in an overhead cable tray, a yellow 480 volt safeguards cable had crossed the metallic barrier. Portions of the cable were now passing over black 480 voit nonsafeguards cables. The inspectors informed the licensee and this condition was correcte The inspectors reviewed the Updated Safety Analysis Report and noted that the cables could be run in the same cable tray as long as the safeguards and nonsafeguards cables were separated by a metallic barrier. Section 8.5 of the Updated Safety Analysis Report also stated that deviation from the standard criteria 1 was acceptable provided an analysis had been completed that justified the deviatio The licensee indicated that the deviation observed by the inspectors was justified in Engineering Analyses EA-FC-90-76 and EA-FC-92-026. The inspectors reviewed these analyses and questioned whether they provided an adequate basis for the licensee's conclusion. The analyses referenced an experiment in which a fault cable (nonsafeguards) was ignited and its effect on a target cable (safeguards) was l observed. The experiment concluded that the target cable only experienced minor smoke damage and still could have performed its designed function. During the experiment referenced by the licensee, the distance between the fault cable and the target cable was 6 inches. However, the distance between the cables observed by i

\.

l

.

-13-l the inspectors was less than 6 inches. Near the conclusion of the inspection period, l the licensee provided the inspectors with an additional analysis to support the position that the safeguards cables could perform the safety functions regardless of a faulted nonsafeguards cable. This item will remain unresolved pending the inspectors' review of the analysis (50-285/96018-05).

l Conclusions I The inspectors identified a potential discrepancy involving cables inside a cable tray in the east switchgear room and questioned whether this condition met the intention of the Updated Safety Analysis Report. This item will remain unresolved pending the inspectors' review of the licensee's analysi E2 Engineering Support of Facilities and Equipment (92903)

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 l review that compares plant practices, procedures, and/or parameters to the Updated l Safety Analysis Report descriptions. While performing the inspections discussed in I this report, the inspectors reviewed the applicable portions of the Updated Safety l Analysis Report that related to the areas inspected. The inspectors verified that the )

Updated Safety Analysis Report wording was consistent with the observed plant i practices, procedures, and/or parameters. One possible discrepancy was discussed I in Section E E8 Miscellaneous Engineering issues (92700)

E8.1 (Closed) LER 50-285/95-001: time delay relays for offsite power low signal found out of tolerance. On March 16,1995, a system engineer reviewing completed ,

offsite power low signal calibration procedures discovered that as-found time delays !

for two of four offsite power low actuation relays were outside the time delay range !

specified in Technical Specification 2-1. The licensee's immediate corrective actions were to reset the out-of-tolerance relays to a value within the range allowed by the calibration procedure and successfully recheck the time delay. Also, the relays were placed on an increased frequency of testing to ensure setpoint stability. A root I cause was performed to determine why the relays were experiencing drif ting setpoints. The vendor supplied information which indicated that the relay remaining i continuously energized for 14 months contributed to the setpoint drift. Based on the root cause analysis, the licensee replaced the relays with a more accurate electronic

)

l relay during the 1996 refueling outage. The inspectors reviewed the licensee's actions and found them to be appropriate.

l The inspectors determined that a violation of Technical Specification Table 2-1,

' Engineered Safety Features System Initiation Instrument Setting Limits, occurred in

that two of four offsite power low signal relays were outside the time delay range l

! _. _

.

.

- 14-

'

, required. This licensee-identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.1 of NRC Enforcement Policy (50-285/96018-06).

E8.2 (Closed) LER 50-285/95-004: discrepancy regarding seismic classification of new .

fuel storage racks. On May 19,1995, as a result of the licensee's efforts to j incorporate the new fuel storage rack design information into the design basis l

documents, a discrepancy was identified. Design records indicated the new fuel rack :

was to be a Seismic Class 11 while the Technical Specification 4.4.1 indicated a l Seismic Class I category. The licensee performed a calculation which determined I that the new fuel storage rack, its anchorage, and the supporting floor had sufficient '

strength to withstand the effects of a safe shutdown earthquake. Based on Engineering Analysis 95-20, the licensee reclassified the new fuel storage racks from Seismic Class in to Seismic Class 1.

-

A review of historical documentation indicated the new fuel storage racks were

'

designed to be Seismic Class ll. Apparently, confusion existed between the l architect / engineer design group and licensee staff who prepared the Technical Specification. Also, the spent fuel storage racks are designed and installed as Seismic Class I structures; therefore, no general concerns were identified with the seismic design of spent fuel storage equipmen !

The root cause of the event was associated with a lack of administrative control during parallel development of the design / installation process and the licensing document process for the new fuel storage rack. The inspectors reviewed the

licensee's actions and found them to be approoriat E8.3 (Closed) LER 50-285/95-007: potential tripping of 480 volt circuit breakers with digital (RMS-9) trip units. On November 8,1995, the plant review committee concluded that the interaction of some safety-related circuit breakers, during certain plant accidents, potentially would constitute a condition outside the design basis of the plant. The licensee performed a safety analysis for operability and performed compensatory actions. The long-term corrective action was to replace the RMS-9 trip unit The trip units were replaced with a Westinghouse Amptector trip device. This work was completed on August 22,1996. The inspectors reviewed the licensee's actions )

'

and found them to be appropriat E8.4 (Closed) Follow-Up Item 50-285/96001-02: instrument air tubing size for the safety injection refueling water tank recirculation valves. This itern was open pending the licensee's identification of all valves that were potentially susceptible to erratic operation based on the size of the instrument air tubing. For the valves identified, the licensee concluded that the valves were operable and capable of performing their design function. Valve f ailures and erratic operations were not attiibuted to instrument air tubing size. The inspectors reviewed engineering change notices for a

<

.

1 i

= j i-15- l l

!

few of the valves that the licensee had identified. The inspectors verified that the licensee had properly evaluated tubing size during equivalency evaluation l l

IV. Plant Support R1 Radiological Protection and Chemistry Controls R1.2 Tour of Radiation Controlled Areas Insoection Scope (71750)

The inspectors performed routine tours of the radiation controlled area and observed radiation work practices of plant personnel, Observations and Findinas I

Throughout the inspection period, the inspectors observed licensee personnel l perform duties in the radiation controlled area. Workers were observed to be obeying l all administrative and regulatory requirements. The inspectors also verified that all I doors required to be locked for the purposes of protecting personnel from radiation exposure were locke On January 30,1997, the inspectors performed confirmatory surveys throughout the radiation controlled area. The areas surveyed by the inspectors had been properly posted to inform workers of radiological condition Conclusions The inspectors concluded that the licensee had properly posted radiation areas throughout the radiation controlled area. Personnel were observed to be complying with all radiation protection requirement R3 Radiation Protection and Chemistry Procedures and Documentation R3.1 Inability to Perform a Chemistrv Procedure inspection Scoce (71750)

The inspectors followed up on the licensee's investigation to determine why a chemistry procedure could not be performed as written, Observations and Findinos While attempting to perform Chemistry Procedure CH-SMP-PA-0005,"High Range Monitoring of Gaseous Effluent Released Via the Auxiliary Building Ventilation Duct Pathway," during a self-assessment on December 4,1996, the licensee discovered l

.

. - . -. - - . -.

t

,

.

-16-that the procedure could not be completed because equipment that should have been located at a designated area had been moved without the knowledge of the chemLtry department. Specifically, Detectc;r E-530, which was required to count a sample of the auxiliary building stack flow, was not in its designated location. This procedure was not in the licensee's preventive maintenance program and was not performed on a specified frec xnc The purpose of the procedure was to outline the action required to monitor effluent release rate of noble gas, iodine, and particulate radionuclides through the auxiliary building ventilation duct in the event that existing instrumentation (RM-063) goes off (

scale or is inaccessible due to abnormally high levels of radiatio Step 6.3 of the procedure directed that personnel access the auxiliary building roof and obtain the equipment needed to sample the auxiliary building stack. The equipment had been stored in a designated storage area on the auxiliary building roof. During the self-assessment, Detector E-530 was not in the storage area.

l

'

The inspectors asked how long the equipment had been moved from its designated location. Chemistry personnel determined that the detector was moved by radiation protection personnel about March 1995. Chemistry personnel stated that the ( detector had been moved to a storage cabinet located at the base of the ladder l leading to the radioactive waste building roof. The detector was moved for industrial ,

l safety reasons and to protect it from harsh environmental conditions. However, l l radiation protection personnel did not inform chemistry personnel that the detector l

had been moved so that all appropriate procedures could be update !

The licensee also determined that a quarterly equipment accountability check was l performed by the emergency preparedness department. The check was performed l using Directive EPT-30, " Equipment inventory - Auxiliary Building Roof." The inspectors reviewed a copy of the checklist and noted that it specified that the detector should be located on the auxiliary building roof in the emergency kit. The licensee determined that, during the last quarter's performance of the procedure, emergency preparedness personnel were aware that the detector was no longer l staged on the auxiliary building roof. The inspectors determined that there was a l I weakness in communications and attention to detail by the emergency preparedness !

l personnel responsible for performing the equipment inventory. The inspectors concluded that, since Procedure EPT-30 was performed on e, quarterly basis, emergency preparedness personnel had numerous opportunities to identify that the equipment was not in the location designated by the directive and to initiate a change to the inventory to correct the equipment location. The inspectors also determined that there was a weakness in the directive in that no guidance was provided to direct emergency preparedness personnel to inform chemistry personnel of any discrepancies with the equipmen Since Procedure CH-SMP-PA-0005 would be performed under postaccident conditions, the inspectors asked what would be the potential radiological effects of

-

. l l-17 ]

I

,

I performing the procedure without the detector being in the specified location. The licensee stated that personnel performing the procedure would potentially be exposed

'

to additional radiation dose. Individuals would obtain the additional potential dose from accessing the roof and determining that the detector was not in the location specified by the procedure. The licensee stated that this additional potential dose could have been avoided if the detector had been identified as not being in the specified location during the quarterly equipment checks and if radiation protection personnel had informed the chemistry department that the detector had been move In discussions with the chemistry manager, he indicated that the procedure was being updated to reflect the new location of the detecto c. Conclusions l

The inability to perform a chemistry procedure was caused by a lack of '

communication between radiation protection, chemistry, and emergency j preparedness personnel. Radiation protection personnel failed to inform chemistry )

personnel that equipment needed to perform the procedure had been moved. There l was no guidance that directed emergency preparedness personnel to inform the I chemistry department that the detector needed to count the auxiliary building stack I sample was not it its designated location. The licensee concluded that, due to the f ailure to communicate, personnel could have potentially been exposed to additional I radiation dose while attempting to perform the procedure in postaccident conditions. ,

1 VI. Manaaement Meetinas l X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management l at the conclusion of the inspection on February 11,1997. The Itcensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie l

- - . .

- . .

l

.

~

!

e ATTACHMENT l SUPPLEMENTAL INFORMATION l PARTIAL LIST OF PERSONS CONTACTED Licensee R. Andrews, Division Manager, Nuclear Services G. Bishop, Assistant Plant Manager R. Connor, Manager, Training D. Dryden, Station Licensing Engineer H. Faulhaber, Manager, Maintenance S. Gambhir, Division Manager, Production Engineering J. Herman, Manager, Outage Management R. Jaworski, Manager, Design Engineering, Nuclear i E. Matzke, Station Licensing Engineer R. Phelps, Manager, Station Engineering H. Sefick Manager, Security Services R. Short, Manager, Operations M. Tesar, Manager, Corrective Action Group D. Trausch, Manager, Nuclear Safety Review Group

. NRC

.

V. Gaddy, Resident inspector l W. Walker, Senior Resident inspector i

f

'

INSPECTION PROCEDURES USED

!

.

IP 37551: Onsite Engineering ,

'

IP 61726: Surveillance Observations l

, IP 62703: Maintenance Observations l l lP 71707: Plant Operations I IP 71750: Plant Support Activities j IP 92700: Onsite LER Review j IP 92901: Followup - Operations i IP 92902: Followup - Maintenance

'

!

IP 92903: Followup - Engineering IP 92904: Followup - Plant Support

,

ITEMS OPENED AND CLOSED Opened

,

j 50-285/96018-02 URI failing to follow configuration change control procedure for

! replacing springs and spiral pins on mainsteam line radiation i monitor isolation valves (Section M1.2).

I i

L

.

e

. - . _ __ _ -

,

I

,

-2- ,

i failing to follow the configuration change control procedure when installing an actuator cylinder on a component cooling water outlet valve (Section M1.3)

l failing to follow the configuration change control procedure )

for installation of a gasket on the safety injection and

refueling water tank vent (Section M1.4)

50-285/96018-03 URI documentation of testing on Valves HCV-921 and HCV-922 (Section M1.2)

50-285/96018-04 URI failing to install flinger ring on component cooling water pump

!

as required by procedure (Section M1.5) l 50-285/96018-05 URI portions of 480 volt safeguards cable passing over nonsafeguards cable (Section E1.1).

i Closed 50-285/95005-00 LER plant trip due to operator error during diverse scram system !

testing (Section 08.1)

50-285/95006-00 LER inoperability of Diesel Generator 2 discovered following a reactor trip (Section 08.2)

50-285/96016-02 URI reportability determinations (Section 08.3)

50-285/95024-01 VIO failure to follow maintenance procedures when performing testing on the raw water / component cooling water heat exchangers (Section 08.4)

50-285/95004-01 VIO failure to follow procedure when performing fuel movement (Section 08.5)

50-285/96001-03 IFl f ailure of incore detectors (Section M8.1)

50-285/95001-00 LER time delay relays for offsite power low signal found out of tolerance (Section E8.1)

50-285/95004-00 LER discrepancy regarding seismic classification of new fuel storage racks (Section E8.2)

50-285/95007-00 LER potential tripping of 480 volt circuit breakers with digital (RMS-9) trip units (Section E8.3)

. _ . _ _ _ . . . . _ _

,

j.

Va u*

a

.,

.

-3-

l

-

!

-

50-285/96001-02 IFl instrument air tubing size for the safety injection refueling j water tank recirculation valves (Section E8.4)

1 Items Opened and Closed

! 50-285/96018-01 . NCV reportability determinations (Section 08.3)

$ 50-285/96018-06 NCV time delay relays for offsite power low signal out of tolerance -

(Section E8.1)

l

l

!

!

)