IR 05000313/1993006
| ML20046C528 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/03/1993 |
| From: | Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20046C522 | List: |
| References | |
| 50-313-93-06, 50-313-93-6, 50-368-93-06, 50-368-93-6, NUDOCS 9308110107 | |
| Download: ML20046C528 (26) | |
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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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Inspection Report:
50-313/93-06 50-368/93-06
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Operating Licenses: DPR-51 NPF-6
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Licensee:
Entergy Operations, Inc.
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Route 3, Box 137G Russellville, Arkansas 72801 t
Facility Name: Arkansas Nuclear One (ANO), Units 1 and 2 Inspection At:
Russellville, Arkansas Inspection Conducted: May 30 through July 10, 1993 Inspectors:
L. Smith, Senior Resident Inspector
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S. Campbell, Resident Inspector
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A. Gaines, Acting Resident Inspector
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Accompanying personnel:
K. Weaver, Engineering Aide N Nb
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Approved:
fW M 6/,o/93 T.' F.'Stetka, Chief, Project Section D Dat6 Inspection Summary Areas Inspected (Units 1 and 2): This routine resident inspection addressed onsite event followup, operational safety verification an engineered safety
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feature system walkdown, observation of monthly, maintenance activities,
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observation of bimonthly surveillance activities, observation of solid
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radioactive waste management and transportation of radioactive materials, control of radioactive materials and areas, and onsite followup of licensee.
event reports.
Results (Unit I):
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The control board reactor operator response during the plant runback to
40 percent power was determined to be appropriate given the information-
available (Section 2.1).
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9308110107 930805 PDR.ADOCK 05000313 G
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An inspection followup item was opened to further evaluate the-variation
of indicated Unit 1 pressurizer level during the plant runback to ensure adequate guidance is provided to the operators regarding future similar events (Section 2.1).
An unresolved item was opened to evaluate the licensee's commitments
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regarding manually cycling the Unit 1 pres.urizer electromatic relief valve to prevent an overpressure reactor trip (Section 2.1).
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The inspectors identified a mispositioned locked throttle valve in the
emergency feedwater pump bearing cooling return line. During this
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inspection period, the licensee also identified other cases where valves were misaligned. The failure to maintain control of valve position was considered to be a violation of Technical Specification 6.8.1 (Section 3.1).
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The inspectors identified a maintenance worker within a
Level I housekeeping area that had not properly secured his badge consistent with the recommendations of the licensee's housekeeping
procedure (Section 3.2).
All valves in the Unit I decay heat removal system were determined by
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the inspectors to be correctly positioned (Section 4.1).
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An inspection followup item was opened to further evaluate the
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causes and corrective actions for the failure of a waste gas decay a
tank rupture disk (Section 5.4).
j Results (Unit 2):
i The licensee evaluation of the Palo Verde steam generator tube rupture e
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event was thorough and resulted in appropriate corrective actions (Section 3.5).
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The procedure for calibrating the margin to saturation monitor contained
incorrect terminal references and required wirelifts without giving
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sufficient detail to ensure the wires could be lifted and restored
correctly (Section 6.2).
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Results (Units 1 & 2):
Switchyard activities were well controlled during a modification to the
auto transformer protective relaying (Section 3.3).
All observed maintenance activities were performed well (Section 5).
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The implementation of the solid radioactive waste management and
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transportation of radioactive materials program was excellent (Section 7).
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-3-The licensee identified a problem with the unrestricted release of
contaminated tools and equipment from the radiological controlled i
area which resulted in a noncited violation (Section 8).
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Summary of Inspection Findings:
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Inspection Followup Item 313/9306-01 was opened (Section 2.1).
e Unresolved Item 313/9306-02 was opened (Section 2.1).
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Violation 313/9306-03 was opened (Section 3.1).
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Inspection Followup Item 313/9306-04 was opened (Section 5.4).
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A noncited violation was identified (Section 8).
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Licensee Event Reports 368/91-005, 313/91-012, 313/91-013, and
368/91-018 were closed (Section 9).
Attachment:
Persons Contacted and Exit Meeting
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DETAILS l
1 PLANT STATUS j
1.1 Unit 1
At the beginning of the inspection period, Unit I was at 100 percent power.
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On June 10 the unit reduced power to approximately 98 percent due to a loss of j
condenser vacuum. The unit returned to 100 percent powar the same day. On June 13, the unit experienced an integrated control system automatic power runback to 40 percent due to a main feedwater pump trip. The unit returned to l
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100 percent power on June 14.
On June 30 the unit commenced a power reduction to 60 percent power due to feedwater heater tube leakage. The unit returned
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to 100 percent power on June 31. At the end of the inspection period, the unit was at 100 percent power.
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1.2 Unit 2
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The unit remained at or near 100 percent power throughout the inspection period.
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2 ONSITE EVENT FOLLOWUP (93702)
2.1 Unit 1 - Loss of Main feedwater Pump (MFWP) P-1A Resulting in a Plant
Runback
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t On June 13 MFWP P-1A tripped due to loss of lube oil pressure resulting in an i
integrated control system automatic power runback to 40 percent. The. _
i inspectors responded to the site following the transient.
Posttransient
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documentation and the applicable procedures were reviewed.
Interviews were i
conducted with the operating crews and with plant management.
The licensee determined that Lube Oil Pump P-26A tripped on thermal overload.
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The Backup Lube Oil Pumps P-27A and P-28A auto started, but did not bring the
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lube oil pressure up soon enough to prevent the trip of MFWP P-1A.
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the loss of one MFWP, the operators carried out the actions of abnormal
operating Procedure 1203.27, " Loss of Steam Generator Feed." The integrated t
control system master was in track and the plant ranback in power. During the transient, the pressurizer spray valves automatically cycled. The j
Electromatic Relief Valve PSV-1000 was manually opened by the operators three l
times for approximately 3 seconds each to prevent a high pressure reactor j
trip. The plant stabilized at 360 MWe which was approximately 40 percent
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reactor power.
l While monitoring pressurizer pressure, the reactor control board operator (CB0) noticed an increasing pressurizer level on Pressurizer Level r
Recorder LRS-1001; however, the level was still less than 290 inches.
Emergency Operating Procedure 1202.01, " Reactor Trip," required a manual l
reactor trip when pressurizer level exceeded 290 inches.
After cycling the
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electromatic relief valve for the third time, the reactor CB0 noticed that the j
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pressurizer level was 300 inches and was trending down after reaching a peak l
level of 305 inches. The operator realized that a reactor trip was required when the pressurizer level exceeded 290 inches and brought this fact to the
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immediate attention of the control room supervisor, the turbine CB0, and the i
extra CB0 on shift.
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As trained, these operators immediately checked the other available i
instruments. The redundant Level Indicator LIS-1002 indicated a level of
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278 inches.
Reactor coolant system average temperature, T,,,, was checked and found to be at 591.5*F.
Based on this T.., the level was estimated by the
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operators to be 282 inches.
In addition, the safety parameter display screen
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indicated a pressurizer level of 280 inches. The operating crew believed the level recorded on Level Recorder LRS-1001 was not valid and noted the trending down on all indicators.
Based upon this information, the operating crew
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decided not to trip the reactor.
The level exceeded the 290 inch-mark on
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Level Recorder LRS-1001 for approximately 20 seconds.
j The pressurizer had two independent level detectors. Level Recorder LRS-1001, i
Level Indicator LIS-1002, and the safety parameter display system were
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displaying temperature compensated level information. The temperature compensation inputs were derived from two thermowells that were located at different elevations. The calculation of the temperature compensation was performed separately for each read-out device. The lower elevation thermowell fed the compensating temperature to the Level Indicator LIS-1002 loop and to
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the safety parameter display value which was selected. The approximately 20-inch level difference was apparently due to the effects of difference in
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temperature compensation caused by an inrush of relatively cool reactor
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coolant.
Further, T,.. was higher than usual for similar previous transients. The licensee determir.ed that a leaking high pressure feedwater heater was the
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predominate contributor to the elevated T,..
The higher T.. apparently resulted in a greater thermal expansion and a larger inrush of reactor coolant into the pressurizer. As a result, the effects of differing temperature
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compensation, due to elevation differences, had not been seen before. The differences in temperature compensation were not modeled on the simulator.
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The licensee provided information about the transient to all the operating crews.
Condition Report 1-93-0174 was initiated to document the licensee's evaluation of the effects of the difference in temperature compensation and the effects of the elevated T.. on pressurizer level indication. The licensee
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planned to incorporate appropriate lessons learned into their requalification program.
Inspection Followup Item 313/9306-01 will be opened to evaluate the
lessons learned to ensure adequate guidance is given to the operators.
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l The appropriateness of manually opening the electromatic relief valve to prevent a reactor trip on overpressure is being evaluated by the NRC. There is concern that such action may be inconsistent with the expectations regarding the implementation of the commitments to NUREG 0737, " Clarification of TMI Action Plan Requirements," Items II.K.3.1 and II.K.3.2.
This issue is
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evaluation.
i 2.2 Conclusions The operator response to automatic-plant runback transient was appropriate.
The plant transient response investigation was excellent.
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3 OPERATIONAL SAFETY VERIFICATION (71707)
3.1 Unit 1 - Control of Valve Alignments On June 8 during routine tours of the plant, the inspectors discovered Bearing Cooling Return Valve CS-1197 to Emergency Feedwater (EFW) Pump P-7B throttled 2-1/2 turns open.
Procedure 1102.001, Revision 51, " Plant Preheatup and Precritical Checklist," Attachment E, Section 1, " Category E Valve Position-
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Verification," required the valve to be in the locked full open position,
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which was six turns open. A waste control operator (WCO), at the request of the inspectors, confirmed the inappropriate position of the valve and
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Condition Report 1-93-0166 was initiated.
The licensee investigated the mispositioned valve and determined that on March 14, a WC0 and a CB0 erroneously mispositioned and independently verified Valve CS-1197 in the throttled position, as a result of failure to utilize self-verification techniques. After the WC0 completed a monthly Category "E" valve position verification, he realized that two individuals were required for independent verification of throttle valves.
Sheet 11 of 19 of Attachment E required that Bearing Cooling Water Supply Isolation Valves CS-1013 and CS-1017 be locked 1-1/2 and 2-1/2 turns throttled open, respectively. The WC0 and CB0 returned to the EFW pump room to reperform l
position verification of Valves CS-1013 and CS-1017 but erroneously selected Bearing Cooling Water Return Valves CS-1199 and CS-1197 instead. The WC0 repositioned Valve CS-Il99 l-1/2 turns open and CS-1197 2-1/2 turns open. The j
full-open position for Valve CS-1199 was 1-1/2 turns open and, therefore, this
'i valve was fully opened. However, the full-open position for Valve CS-1197 was six turns open. This valve remained throttled 2-1/2 turns open through four
'j succeeding monthly Category "E" valve position verifications.
After the i
mispositioned valve was identified, the valve was unlocked, fully opened,.
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relocked, and the valve position was independently verified.
The licensee's operability determination, which conservatively assumed a
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linear flow through the valve, concluded that a bearing cooling flow i
65 percent of normal flow was sufficient for pump operation.
The condition
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report did not include the basis for the engineering judgement. The inspectors discussed the operability determination with a systems engineer who
provided a reasonable explanation for the basis. The failure to document the basis for the engineering judgement in the condition report was a weakness.
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During the inspection period, the licensee identified other misaligned valves.
l On June 3, DG1 Air Dryer M-300A/B Bypass Valve F0-92A1, around an air dryer
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for the emergency diesel generator was found open.
Procedure 1104.036, i
Revision 32, Permanent Change 1, " Emergency Diesel Generator Operation,"
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Attachment A, "DG1 Valve Lineup," required that this valve be normally closed.
The valve was closed and Condition Report 1-93-0159 was initiated.
No unusual
amount of moisture was discovered during a blowdown of the air receiver tanks.
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Therefore, the licensee determined that the operability of emergency diesel generator air start system was not affected.
On June 24, Core Flood Makeup Tank Inlet Valves MU-51 and MU-52 were not closed following the completion of
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boric acid pump testing at approximately 5:30 a.m.
Step 2.9 of Procedure 1104.003, Revision 26, Permanent Change 1, " Chemical Addition,"
Supplement 2, " Boric Acid Pump P-39A Test," required that these valves be closed.
The misalignment resulted in degraded emergency boration capability
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and an unplanned entry into Technical Specification 3.2.1.3 which allowed the i
system to be out of service a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The valves were closed at
approximately 12:25 p.m. the same day and Condition Report 1-93-0191 was initiated.
Failure to maintain the correct alignment of Valves CS-1197, FO-92A1, MU-51,
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and MU-52 as required by plant procedures was considered to be a violation of
Technical Specification 6.8.1 (313/9306-03).
i 3.2 Unit 1 - Spent Fuel Pool Fuel Handling Bridge
On June 11 while repositioning fuel assemblies in the spent fuel pool for the
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upcoming refueling outage, a fuel assembly could not be ungrappled.
The inspectors went to the spent fuel pool area to observe the condition. The inspectors were informed that the fuel assembly was fully lowered in the X-43 position and that there was a hole in the upper hydraulic line which prevented the ungrappling of the fuel assembly. The licensee decided to tape the hole with duct tape to see if that would hold long enough for the fuel
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assembly to be ungrappled. The taping of the hole proved to be unsuccessful
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and the licensee had to manually ungrapple the fuel assembly. The fuel handling bridge was then parked and locked. The two upper hydraulic hoses
were subsequently replaced and the fuel handling bridge returned to an
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operable condition on June 14.
The inspectors observed personnel exiting the area surrounding the spent fuel
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pool. The spent fuel pool was bounded by two barriers; the inner barrier was a contaminated area barrier with step-off pads, and the outer. barrier
indicated that the enclosed area was a Level I housekeeping area. The outer barrier was located approximately 10 feet from the inner barrier. The inspectors noted that when one individual was taking off his anticontamination-
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clothing at the step-off pad, his security badge and thermoluminescent
~ dosimeter (TLD) were clipped to the inner clothing.
The inspectors' questioned licensee personnel, covering the fuel handling bridge work, if the clipping of
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the badge 'and TLD to the inner clothing was appropriate for the Level I
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housekeeping area.
The licensee's response was that such items should be on a i
lanyard or taped down.
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The individual had been performing work on communications equipment located approximately 6 feet from the spent fuel pool. The inspectors determined that when the individual entered the Level I housekeeping area, the badge and TLD were secured by being inside the anticontamination clothing. However, due to the different locations of the two barriers, when the individual took off the L
anticontamination clothing at the step-off pads, he was still in a Level I housekeeping area, and the badge and TLD were not secured.
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A review by the inspectors of Section 6.2.4 of Procedure 1000.018, e
" Housekeeping," Revision 19, for a Level I housekeeping area specified
controls to prevent foreign material from being introduced into systems,
structures, or components.
These controls should include securing all items on a person (i.e., security badge, glasses, pens) with tape or safety straps.
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Further discussions with the licensee's staff indicated that their expectation in the above case would be that the badge and TLD be taped down. The inspectors informed the licensee that the. individual's failure to secure his
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badge and TLD appeared to be a weakness. The licensee stated that the procedure would be revised to clarify the intent that items shall be secured
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while in a Level I housekeeping area.
j 3.3 Unit 2 - Operations Control of Switchyard Activities l
On July 8 the inspectors reviewed the control of switchyard activities. A modification to the fault detection equipment for the autotransformer was being installed.
Prior to the installation, a plant impact statement and a
component impact statement were prepared in accordance with Procedure 1015.33,
"ANO Switchyard and Transformer Yard Controls." - The plant impact statement
plus a projected schedule for the work activities was supplied to the shift
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superintendent. The plant impact statement required, among other things, that no other activities be in progress in the switchyard during the installation of the additional fault detection equipment.
Further, the procedure required
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that high risk activities such as control rod drive mechanism breaker trip i
testing not be performed. The control rod drive mechanism breaker trip testing was appropriately scheduled on July IO,-following the completion of the autotransformer fault detection modification.
Prior to performance, the licensee carefully evaluated the switchyard activities and effectively communicated relevant information to the operating staff.
3.4 Unit 2 - Temporary Modification Control j
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The_ inspectors reviewed 10 temporary modification packages to ensure the affected control room drawings had been identified and marked. This was to I
ensure the ' package had been logged into the control room log and adequate postinstallation testing was specified. No problems were identified.
A recent program improvement which-required a periodic temporary modification audit by the systems engineer was effective.
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3.5 Unit 2 - Operations Review of Palo Verde Steam Generator Tube Rupture Event On March 14, 1993, a steam generator tube rupture event occurred at the
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Palo Verde plant.
The event was not correctly diagnosed by the licensee for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, even though a steam generator tube rupture event was suspected, because the event did not meet the entry conditions for the site's emergency operating procedures.
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Because of the steam generator tube leakage event that occurred at ANO during 1992 and similarities between the units, the licensee re-examined their emergency operating procedures to ensure a steam generator tube rupture event would be effectively diagnosed at AN0 under similar conditions.
Personnel were sent from ANO to Palo Verde to identify operating philosophy differences.
The licensee provided the inspectors with a summary comparison for the differences in emergency operating procedure use and the differences in operating philosophy.
The inspectors interviewed the operations manager and several of his staff members. They stated the Palo Verde event and associated lessons learned were discussed during shift meetings with the operating crews.
Lessons re-emphasized to the operating crews included:
(1) the new N" radiation monitors are at power monitors which will only alarm while at power and subsequently clear following a reactor trip, (2) the blowdown monitors will not indicate accurately, if the isolation valves have closed because of a safety injection actuation signal or a containment isolation actuation signal, and (3) reliance on more than one indication to formulate a decision as well as a questioning attitude is as important as procedural compliance.
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licensee expected to conduct additional training during the next
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requalification cycle.
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Emergency Operating Procedure 2202.01, " Standard Post Trip Actions," was reviewed by the inspectors to confirm the validity of the licensee's conclusions regarding emergency operating procedure usage at ANO. The standard posttrip actions required that relevant trends be evaluated rather than relying solely on a particular value and the diagnosis flow chart provided direction for any indication of secondary radiation. Therefore, it appears that entry into the ANO steam generator tube rupture emergency
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operating procedure would have been more timely thus providing earlier onset
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of mitigating activities.
t 3.6 Conclusions Valve configuration control was not well maintained on Unit I resulting in a violation of Technical Specification 6.8.1.
The licensee improved their housekeeping procedure following the identification of a weakness by the inspectors.
The control of switchyard activities was effectively
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proceduralized and implemented. Temporary modification control on Unit 2 was effective. The proactive response to the Palo Verde steam generator tube rupture event was prudent and was viewed as a strength.
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4 ENGINEERED SAFETY FEATURE SYSTEM WALKDOWN (71710)
4.1 Unit 1 - Decay Heat Removal System Walkdown The inspectors performed a walkdown of the decay heat removal system to verify valve alignment and system operability.
Piping and Instrumentation Diagram M-232, Revisioi 75, " Decay Heat Removal System," was used as a guide for valve alignment. All valves were identified and aligned per the diagram, and no gross packing leaks associated with the applicable valves were noted.
Both Decay Heat Removal / Low Pressure Injection Pumps P-34A and P-34B had adequate inboard and outboard bearing lube oil levels.
No oil leakages were
identified. No loose hangers or supports associated with the system were identified. Housekeeping was good and no flammable materials were identified in the areas.
The inspectors reviewed Procedure 1104.004, Revision 54, " Decay Heat Removal Operating Procedure," and used Attachment A, " Decay Heat Removal System Valve Lineup for Engineered Safeguards Actuation (Emergency Standby)," to check the lineup indicated in the procedure with the piping and instrumentation diagram.
All valves identified in the procedure were noted to be aligned as shown in the diagram. Subsequent control room tours by the inspectors verified that the control panel valve indications were consistent with the proper valve al ignment.
4.2 Conclusions The walkdown indicated that all valves were aligned according to the controlling Procedure 1104.004.
5 MONTHLY MAINTENANCE OBSERVATION (62703)
5.1 Unit 1 - Inspection and Lubrication of Make-Up Pump P-36A (JO 00888439)
On June 4 the inspectors observed the licensee's performance of Procedure 1411.010, Revision 4, " Unit 1 Primary Heat Make-Up Pumps Lubrication and Inspection." The licensee replaced the pump coupling grease with
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appropriate grease specified in the procedure and the vendor catalog. The
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procedure was cross-referenced with the vendor catalog by the inspectors and both documents were in agreement in terms of work instructions. Difficulty was experienced in removing one lube plug because the plug was tightened to 50 ft-lbs during the previous lubrication of the coupling. Numerous
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occurrences of stripping and rounding of lube plug allen flats caused the licensee to issue Plant Change I to the procedure to lower the lube plug
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torque value to " snug tight." Additionally, Technical Manual B580.0020,
" Byron Jackson Type DVMS 3x4x8 Nine Stage Primary Makeup Pump," which specified a torque value of 50 ft-lbs, was changed by the licensee with i
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I-11-permission of the vendor to " snug tight." However, the mechanic identified to
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the inspectors that Step 3.6.6 still specified a lube plug torque value of i
50 ft-lbs. The step was inadvertently missed during the first plant change.
The mechanic requested and obtained Plant Change 2 to correct the torque value.
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5.2 Unit 1 - Operational Verification of Sodium Hydroxide Discharge Valve CV 1617 Utilizing the Valve Operating and Test Evaluation System (VOTES) (JO 008878767)
On June 15, 1993, Sodium Hydroxide Discharge Valve CV-1617 in Train B of the Reactor Building Spray System was tested using the. VOTES. The electricians performed the test in accordance with Procedure 1403.038, Revision 7, " Units 1 and 2 MOV Testing and Maintenance of Limitorque SMB-000 actuators,"
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Supplement 11.
The spring. pack and the V0TES testing results were acceptable.
t'o discrepancies were noted between the Vendor Technical Manual L200.001 and
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the procedure.
The subsequent stroke time testing of the valve was successful.
5.3 Unit 1 - Differential Pressure Testino of Loop II Service Water to Intermediate Cooling Water Cooler Supply Valve CV-3811 (JO 00861000)
j On June 21 V0TES was used to perform differential pressure diagnostics testing-
of Intermediate Cooling Water Supply Valve CV-3811. The electrical
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t maintenance was performed in accordance with Procedure 1403.038, Revision 7,
" Units 1 and 2 MOV Testing / Maintenance of SMB-000 Actuators," Supplement 13.
The as-found data for the motor operator spring pack was determined to be acceptable under static conditions and during flow conditions. No
discrepancies were identified.
l 5.4 Unit 1 - Replacement of Waste Gas Decay Tank T-18B Rupture Disk
(JO 00888929)
On February 11, 1993, a condition report was written to document the failure of Waste Gas Decay Tank T-18A rupture disc. The rupture disk failed as a result of overpressurizing the tank above the rupture disc rated pressure of 125 psig.
Following the completion of an 18-month calibration of the gaseous radwaste system in accordance with JO 00888063 and Procedure 1304.132, Revision 4, t
" Unit 1 Gaseous Radwaste Instrumentation Calibration," on February 11, instrumentation and control (I&C) technicians notified the control room that the test was completed.
The control room operators misinterpreted the
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completion of the test to mean that the system was in normal alignment which
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had Gaseous Waste Discharge Valve CV-4830 open. The I&C technicians continued i
working to restore the system and, as a result, closed Valve CV-4830.
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Operating Procedure 1104.022, Revision 26, " Gaseous Radwaste System," required Waste Gas Decay Tank T-18B to be removed from service when the tank reached a i
pressure of 88 psig. When the operators received a high pressure alarm (set at 110 psig) on this tank in the control room, they dispatched the WC0 to isolate the tank; however, the rupture disk ruptured before the tank could be isolated.
No unplanned gaseous releases occurred as a result of the rupture disk failure.
l To prevent recurrence of this event, the licensee planned to revise Procedure 1304.132, " Unit 1 Gaseous Radwaste Instrumentation Calibration," to allow operator discretion regarding the final valve position.
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the licensee is performing an evaluation to determine whether a reduction in the high pressure alarm setpoint from 110 psig to 85 psig should be made to
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provide sufficient time for operator response.
Further review of this event
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including the reason that this event had not occurred during past calibrations
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is considered to be Inspection Followup Item (313/9306-04).
- j On June 21, 1993, the inspectors observed the replacement of the Waste Gas
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Tank T-18B rupture disk.
The waste gas decay tank was depressurized and
isolated. Health physics personnel surveyed the area and established a
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contamination boundary prior to removing the rupture disk.
No instructions
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were provided in the J0 package on how the rupture disk should be oriented;
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however, a tag on the replacement disk specified the correct orientation, and
the mechanics installed the disk correctly and torqued the flange nuts using a
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calibrated torque wrench.
The mechanics used good radiological work practices
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5.5 Unit 1 - Replacement of Post Accident Sampling System Valve CV-5963
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(JO 00895640)
On June 24 qualified welders ground out the existing welds on Valve CV-5963 i
and removed the leaking valve from the post accident sampling system. A new
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like-for-like replacement valve was installed. A qualified firewatch was
present and foreign material exclusion was maintained during the welding i
activity. A quality control inspector inspected the final welds. The
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inspectors observed the performance of the activity in accordance with the
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JO instructions.
I 5.6 Unit 2 - Inspection and Cleanino of low Pressure Safety Injection i
Pump 2P-60A Seal Cooler 2E-52A (JO 00896053)
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A decreased service water flow trend to Low Pressure Safety Injection Pump 2P-60A Seal Cooler 2E-52A prompted the licensee to clean the seal cooler i
and the strainer upstream of the cooler. On June 30 the inspectors observed
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the licensee disassemble, inspect, and clean leafy debris from the shell side of the cooler.
The strainer upstream of the cooler contained silt deposits.
The seal cooler was inspected, reassembled, and appropriately torqued. A
7 percent increase in service waterflow through the cooler was measured during i
postmaintenance testing.
The inspectors reviewed the new revision to l
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Table 9.2-1, " Service Water System flow Rates," of the Unit 2 Safety Analysis Report and confirmed that minimum flow rate requirement was met.
5.7 Conclusions All maintenance activities throughout the inspection period were performed well. The licensee provided changes to procedures to address the recurring problem of stripped allen flats on the make-up pump coupling grease fittings.
6 BIMONTHLY SURVEILLANCE OBSERVATION (61726)
6.1 Unit 1 - Testing of Reactor Protection System Train B Control Rod Drive Breaker (JO 00889382)
On July 2,1993, the inspectors observed I&C technicians perform an "as found" i
trip test on Control Rod Drive Train B breaker per Procedure 1304.126, Revision 6, " Unit 1 RPS - B/CRD Breaker Trip Test." The I&C technicians
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performed the procedure in sequence as required by the procedure.
The breaker was reset and electrical maintenance personnel replaced the breaker with a spare breaker that was cleaned and inspected under J0 00896494. The breaker replacement was performed in accordance with Supplement II of Procedure 1405.017, Revision 10, " Unit 1 Reactor Trip Breaker Inspection."
The I&C technicians performed the remainder of the trip test procedure for the newly installed breaker. The time response test results were within procedural requirements.
6.2 Unit 2 - Calibration of Margin to Saturation Instrumentation
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(JO 00896241)
On July 9, 1993, the inspectors observed the performance of portions of
Procedure 2304.078, Revision 11, Permanent Change 1, " Margin to Saturation
Instrumentation Calibration."
The procedure did not provide for independent verification that resistance
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temperature detector leads were appropriately disconnected and reterminated.
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The I&C technicians appropriately used a lifted lead Log Sheet Form 1025.003B.
The inspectors noted that Step 8.8.8 of the procedure incorrectly referred to j
Terminal Block TB4-10, Points 17 and 18 prior to the performance of that step.
The I&C technicians stopped worked, contacted their supervision, and initiated
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a procedure revision to provide the correct terminal block points prior to
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resuming work. The inspectors reviewed this issue and determined that this
procedure was last performed under Revision 10. Apparently, a typographical error regarding these terminal blocks occurred during the revision process i
thus resulting in the incorrect procedure step. Following the additional procedure revision, the surveillance test was successfully completed on July 10.
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On July 10, the inspectors observed the exercising of the control rod drive mechanism (CRDM) in accordance with Procedure 1105.009, Supplement 2,
" Exercising CRDM's Above Cold Shutdown." The operator carefully performed the procedure. The procedure did not give step-by-step instructions in all cases.
However, the level of detail was appropriate given the operator's level of training and the available supervision.
No problems were identified.
6.4 Conclusions The observed surveillance testing was performed well and in accordance with Technical Specification requirements.
However, Procedure 2304.078, " Margin to Saturation Instrumentation Calibration," contained an incorrect terminal block reference and did not provide for documentation of independent verification for required wire lifts. This was considered to be a weakness.
7 SOLID RADI0 ACTIVE WASTE MANAGEMENT AND TRANSPORTATION OF RADI0 ACTIVE MATERIALS (86750)
The inspectors reviewed the licensee's radioactive material transportation program to determine agreement with the commitments made in response to NRC Bulletin 79-19; compliance with the requirements of 10 CFR Parts 20, 30, and 71; and 49 CFR Parts 171 through 189. The inspectors also reviewed the licensee's program for processing, control, and onsite storage of solid radioactive waste for agreement with the commitments in Chapter 11 of Units 1 and 2 Updated Safety Analysis Reports and compliance with the requirements in Unit 1 Technical Specification 4.29.4 and Unit 2 Technical Specification 3.11.4; and 10 CFR 20.301, 20.311, 61.55, and 61.56.
7.1 Audits and Appraisals Audit and surveillance reports of quality assurance activities performed, l
I since the last NRC inspection of the solid radioactive waste and transportation programs in February 1992, were reviewed for scope, thoroughness of program evaluation, and timely followup of identified deficiencies. Quality Assurance Audit Report QAP-1-92, "Radwaste Management,"
was performed October 20 through December 16, 1992, in accordance with quality assurance procedures and schedules by qualified auditors, who were
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knowledgeable in solid radioactive waste and transportation program j
requirements at nuclear power facilities. The inspectors noted that the audit identified pertinent findings and that prompt corrective actions were taken to correct the findings.
The audit of the solid radioactive waste and transportation programs was of good quality and satisfactory to evaluate the (
licensee's performance of implementing the solid radioactive waste and transportation program.
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-15-The inspectors reviewed the quality assurance surveillances performed during the period February 1992 through May 1993 in the areas related to the performance of the solid radioactive waste and transportation programs.
The quality assurance surveillances were of good quality and satisfactory to evaluate the licensee's performance and provide periodic management oversight.
7.2 Changes The inspectors reviewed the organization, management controls, staffing, and the assignment of solid radioactive waste and transportation program responsibilities for changes. The inspectors noted that there had been no
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major changes in facilities, equipment, programs, and procedures that would j
t have adversely affected the solid radioactive waste management and transportation of radioactive materials programs, since the last inspection.
The inspectors noted the minor organ'zational changes that were made to the i
program. These changes were discussed in NRC Inspection Report 50-313/93-05;
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50-368/93-05.
7.3 Training and Qualifications f
The inspectors reviewed the training and qualification programs for personnel
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i responsible for implementing the solid radioactive waste and transportation of radioactive materials programs..
The inspectors reviewed individual staff l
computerized training records for selected individuals. The licensee's
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training records indicated that the radwaste training course had been
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conducted annually and was last conducted in December 1992. The training i
included Department of Transportation Regulations, 10 CFR Part 61
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and 10 CFR 20.311, and site-specific requirements for low-level waste burial
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sites.
In a March 1993 letter to the NRC, the licensee changed their commitment to Item 6 of NRC Inspection and Enforcement Bulletin 79-19, l
" Packaging of Low-level Radioactive Waste for Transport and Burial."
l Item 6 requi-ed licensees to conduct periodic retraining of personnel involved j
in radioactive waste processing, packaging, and transportation. The licensee j
changed their commitment from an annual to a biennial frequency.
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i The inspectors reviewed lesson plans and discussed training with the I
radioactive waste instructor. The lesson plans were good. The instructor
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indicated that he attends radioactive waste workshops every year.
The j
instructor also stated that, in addition to the annual training provided the i
radioactive waste personnel, they sometimes have vendors teach special topics.
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In April 1993 they had a vendor instruct the personnel on loading of the j
trailers used to transport radioactive waste or materials.
i The inspectors determined that the radioactive waste operations department had
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f an adequate, qualified staff to perform radioactive waste activities.
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7.4 Solid Radioactive Waste Management The inspectors reviewed selected radioactive waste procedures that implemented
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the licensee's solid radioactive waste management program.
The inspectors
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noted that the procedures were adequate for the processing and disposal of
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low-level radioactive waste and met the requirements of the licensee's Technical Specifications.
i The inspectors reviewed the licensee's records for low-level radioactive waste I
shipped since 1990. The following tabulation shows the ' total volume and curie content of the low-level radioactive waste shipped for the period 1990 through j
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December 11, 1992.
Year Volume - Cubic Meters Curie Content l
1990 169.3 14.21 1991 418.0
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1992 175.3 2,605.8
In 1992 approximately 2,560 curies of the curie content that was shipped was
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from the licensee's shipments of spent resins, filter sludges, and evaporator I
bottoms.
In 1990 and 1991, the licensee made an irradiated fuel shipment.
The licensee did a very good job of identifying and shipping for burial the
majority of radioactive waste onsite before July 1, 1993, to preclude interim onsite storage while the uncertainties of future burials are being resolved.
7.5 Radioactive Waste Classification. Waste Characterization, and Shipping i
Reauirements
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The inspectors reviewed the licensee's radioactive waste procedures and found the licensee's program for classification and characterization of radioactive
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waste to meet the requirements of 10 CFR Part 61. The licensee and a contractor laboratory performed radiochemical analyses on samples of various
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radioactive waste types for-the requirements in 10 CFR 61.55 and 61.56. The
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test sample analyses results were used for determination of radwaste classification and isotopic composition of the radwaste sources. The licensee
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- performs isotopic analysis for isotopic characterization on radioactive waste '
i packaged for shipment and burial and employs correlation factors for j
characterization of isotopes not directly identified.
The inspectors reviewed selected radioactive waste shipment manifests and shipping papers that accompanied the licensee's shipments of radioactive
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waste. The inspectors determined that the completed manifests complied with.
I the requirements of 10 CFR 20.311.
7.6 Transportation Activities
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i The inspectors reviewed the licensee's transportation program for shipment of j
radioactive materials and radioactive waste.
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7.6.1 QA Program The licensee has maintained an approved (Approval 0341) quality assurance
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program in accordance with 10 CFR Part 71, Subpart H, for the transportation e
of radioactive materials. The approval expires January 31, 1995.
l 7.6.2 Procurement and Selection of Packages
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The licensee used strong-tight containers for the shipment _of radioactive waste. Of the 161 shipments made in 1992, 75 were laundry shipments in steel containers. Most of the other shipments were in sea / land containers which
contained uncompacted waste that was shipped to a vendor who segregated and repackaged the radioactive waste.
The licensee was on the user's list for all NRC and DOT certified packages used. The licensee maintained current documentation on the manufacturer's design testing, maintenance, and the NRC Certificate of Compliance for all radioactive material packages used by the licensee.
7.6.3 Preparation of Packages for Shipment The inspectors verified that the licensee had procedures and checklists for the preparation of radwaste shipments. These procedures provided for visual inspection of the package prior to filling the container, instructions for closing and sealing the container, marking and labeling requirements, and determination of radiation and contamination limits.
The licensee routinely used a checklist to assure that the procedures were followed and that packages were properly prepared for shipment in accordance with NRC, DOT, state, and burial site requirements.
Discussions with licensee personnel indicated that the individuals involved in the transportation of radioactive waste and materials possessed a working knowledge of the procedures and NRC and DOT regulations pertaining to the preparation of packages for shipment.
7.6.4 Delivery of Completed Packages to Carriers The inspectors verified that the licensee's procedures included the required NRC and DOT regulations. A review of selected records and shipping papers for radioactive waste shipments indicated that the licensee had prepared appropriate manifests and shipping papers in accordance with approved procedures. The. shipping papers included the necessary information to comply with regulatory requirements. The licensee only used exclusive use carriers for all radioactive waste shipments and assured that the following items were in accordance with NRC and 00T regulations and station procedures:
radiation levels were within required limits, transport vehicles were properly placarded, surface contamination on packages did not exceed requirement levels, and blocks and/or braces were in place to prevent damage or shifting of the load during transi.
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7.6.5 Records, Reports, and Notifications
The inspectors reviewed selected records of radioactive waste shipments made by the licensee during 1992.
The licensee's shipments were adequately i
documented to meet NRC and D0T regulations. The licensee maintained records
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of all radioactive waste and materials shipments as required. The records included all shipping documentation, radiation surveys, and required
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notification data.
7.7 Conclusions Excellent audits and surveillances were performed by qualified individuals.
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The audits and surveillances identified pertinent findings and corrective
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actions for the findings were timely. There had been no major changes in facilities, equipment, programs, or procedures.
The radioactive waste operations department had an adequate, well-qualified
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staff to meet staffing requirements. The licensee had maintained a good training program for radwaste personnel.
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The licensee had good implementing procedures for the radioactive waste management program. - The licensee performed an excellent job of identifying and shipping radioactive waste for burial in 1992 and the first half of 1993.
The licensee's low-level radioactive waste disposal program was conducted in accordance with the requirements of 10 CFR 20.311, 61.55, and 61.56.
The licensee maintained good implementing procedures that addressed waste classification and characterization, selection of packages, preparation of packages, and delivery of the completed packages to the carrier.
Individuals responsible for transportation of radioactive waste were knowledgeable of the regulatory requirements and burial site license conditions.
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8 CONTROL OF RADI0 ACTIVE MATERIALS AND CONTAMINATION, SURVEYS, AND
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MONITORING (83750)
On April 14 the licensee's personnel performed a weekly radiation survey of
the maintenance facility tool room. Two slings were found to be contaminated and had fixed contamination levels of 200 and 140 counts-per-minute (cpm).
Additional surveys were performed on April 14, but they did not identify other
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contaminated material. The event was documented by Radiological Information
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Report 93-033.
On May 5 another routine radiation survey of the maintenance facility tool
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room was performed. A sling was found to be contaminated with fixed contamination levels of 2,000 cpm. A followup survey on May 5 was performed of the maintenance facility tool room and five more slings were identified;
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four of the slings had fixed contamination levels of 200 cpm or less, and one
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The results of the surveys were
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documented in Radiological Information Report 93-038 and Condition Report C-93-0058.
The licensee began an investigation of the events and determined that the six slings found May 5 appeared to be in the maintenance facility tool room
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during the April 14 survey. The reason the slings were not discovered, as determined by the licensee, was due to a failure to communicate specific
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information regarding the contaminated slings during shift turnover. During the May 5 incident, there was a direct turnover between shifts which resulted
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in the oncoming shift knowing that there was a problem with the slings.
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Therefore, they surveyed all the slings and found five additional slings, i
The licensee reviewed the tool control process and the radiological control process of tools and equipment and identified numerous weaknesses. The
weaknesses dealt with the logging in and out of tools on the computer, decontaminating tools which sometimes removed the purple paint used to
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identify contaminated tools, removing tools from the decontamination area before they are decontaminated, removal of tools from the tool room when it was not open, use of tools from the maintenance facility tool room in isolated
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radiological controlled areas (RCAs) outside the main controlled access area, mixing clean and contaminated tools tugether in an area like the borated water
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storage tank area which requires clean and contaminated tools to be'used, large volumes of tools and equipment removed from containment at the end of an outage, and slings that are used on a noncontaminated side of a boundary to
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move material to a contaminated area.
The licensee reviewed release pathways through which contaminated equipment or tools could be released from RCAs, survey methodology used during routine surveys, previous radiological information reports, condition reports, and the issue and return history of the contaminated slings. The licensee initiated an action plan to perform a comprehensive radiation survey of the site to
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identify other contaminated equipment, and to control any further release of
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contaminated equipment from the RCAs or from the site. The action plan
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included:
Verifying that vehicles leaving the protected area or warehouse area do
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not contain hand-held tools or equipment that have not had a radiation l
survey performed on them;
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Identifying release pathways of contaminated material from the site and
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high potential locations were contaminated materials could be taken or j
i stored so these areas could be surveyed; and, Restricting the use of only those tools from the maintenance facility
tool room that had been released by health physics and requiring I
equipment issued from the nuclear steam supply system warehouse to have a radiation survey prior to removing the item from the warehouse.
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The high potential locations on site at which contaminated equipment could be
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used or stored included the following:
Inside the Protected Area:
the maintenance facility tool room, Nuclear'
Steam Supply System Warehouse, and the Main Warehouse.
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Outside the Protected Area: the Ritchie Training Center, Q Laydown
Area, Outage Laydown Area, Modification's Fab Shop, Bechtel Tool Trailers, Hickey Warehouse, and all outside warehouses.
The licensee prioritized these areas and began contamination surveys of the equipment in these areas. The surveys consisted of hand frisking individual
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tools and pieces of equipment with a RM-14 survey meter with a HP-210 probe and sometimes using a TCM-2 automatic contamination monitor. The acceptable
limits used for free release of the tools or equipment were as specified in
Section 6.2.8.A of Procedure 1012.020, " Radioactive Material Control,"
Revision 0, which states that material can be " Free Released" from an RCA
providing it does not have inaccessible areas that have the potential of being i
contaminated; and, loose contamination is less than 100 cpm above background
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by smear survey and fixed contamination is less than 100 cpm above background
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by direct frisk, or it does not alarm an automated contamination monitor.
l As of the exit on July 13, the licensee had finished surveying some areas and expected to be finished with all of the surveying by the middle of September.
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The licensee had surveyed over 56,000 items and of these items surveyed the
licensee had identified 30 items, in addition to the slings previously found,
that were contaminated above the free release criteria. One item located in
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the maintenance facility tool room had 3,000 disintegrations per minute loose contamination. This item was a special tool which had last been used in 1983.
l Twenty-five items had fixed-contamination levels of between 100 and 400
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corrected counts per minute.
Four other items had 10,000, 2,600, 1,600,.
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and 1,000 corrected counts per minute fixed contamination levels.
Five of the l
items were located outside the protected area, three at the training center,
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one at the Q laydown area, and one at a fabrication shop.
The licensee determined that under the current system of tool control, there
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was no way to identify how or when a contaminated tool or piece of equipment
was released from the RCA or outside the protected area. _However, from their
review and discussions with personnel, they believe that most of the
contaminated tools are a result of a past inadequate free-release program
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during the 1989-1991 period.
Even though the contaminated tools appear to be i
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a problem from the past, the licensee's review identified areas for improvement in their current program.
In addition to the corrective actions discussed previously, the licensee has performed the following:
. A training evaluation action request was written to establish an annual'
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requirement for reviewing RCA control point operations during the health l
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contractor training.
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i Procedures were revised to clarify the maximum background count rate
while performing a direct frisk survey.
Y A lessons learned summary of the event emphasizing the significance of
the event and the performance expectations in controlling contaminated tools and equipment, and describing future plans for improving control i
of these items was developed.
i Plant outage managers were asked to incorporate information on the e
control of contaminated tools and equipment in their outage handbooks.
l The current tool control system was evaluated and recommendations for
improvements based on the weaknesses identified in the condition report
were evaluated by an independent work group and presented to the i
Unit 2 plant manager.
Improvements in identifying contaminated tools and equipment to ensure
they are not used outside of the RCAs were recommended and evaluated.
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Expectations were provided to health physics technicians and supervisors
on the importance of detailed turnovers when investigating radiological events.
Guidance was provided to health physics personnel for conducting surveys
of areas with a high potential for containing contaminated tools and equipment.
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At the exit meeting on July 13, the general manager, plant operations, i
committed to the following corrective actions:
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Complete the comprehensive site survey of hand tools and equipment to
identify any remaining contaminated items outside RCAs.
Completion date
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for this item is September 15, 1993.
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Establish an annual requirement for reviewing RCA control point
operations during the health physics continuing training program, and during health physic contractor training.
Completion date for.this item is September 15, 1993.
Incorporate training materials into GET/ outage training on how
contaminated tools and equipment are identified and how these items
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could potentially be removed from RCAs without proper monitoring.
Completion date for this item is September 15, 1993.
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-22-Provide site personnel training on this event including lessons learned
and expectations for performance of controlling contaminated tools and equipment.
Completion date for this item is September 15, 1993.
Develop and implement improvements to the existing process for visibly
identifying radioactive tools and equipment at Arkansas Nuclear One.
Completion date for this item is November 1, 1993.
Units 1 and 2 Technical Specification 6.8.1. require, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33 which, in turn, recommends radiation surveys be covered by procedures.
Section 6.2.8.A of Procedure 1012.020, " Radioactive Material Control," Revision 0, states that material can be " Free Released" from an RCA providing it does not have inaccessible areas that have the potential of being contaminated; and, loose contamination is less than 100 cpm above background by smear survey and fixed contamination is less than 100 cpm above background by direct frisk, or it does not alarm an automated contamination monitor. The free releasing of items that had loose contamination or fixed contamination greater than 100 cpm above background and would have alarmed an automated contamination monitor was viewed as a violation of Technical Specification 6.8.1.a.
This violation will not be subject to enforcement action because the licensee's efforts in identifying and correcting the violation meet the criteria specified in Section VII.B.2 of Appendix C to 10 CFR Part 2.
8.1 Conclusions The licensee identified a problem with contaminated tools and equipment that were released from RCAs with contamination levels above their free release criteria. The problem appeared to be due to a poor, free-release program in the past. The licensee took prompt and effective corrective actions to correct the problem. Therefore, the violation identified for releasing material from a RCA with contamination levels above the limits specified in the licensee's procedure for free release was not cited.
9 ONSITE FOLLOWUP OF LICENSEE EVENT REPORTS (92700)
9.1 (Closed) Licensee Event Report 368/91-005:
" Inadvertent Actuation Of Protective Relay Due To Vibration Results In Trip Of Reactor Coolant Pump Motor Breaker and Subsequent Automatic Reactor Trip" This licensee event report involved the actuation of a motor phase differential protective relay for Reactor Coolant Pump 2P-32B which resulted
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l in an automatic reactor trip due to loss of power to the reactor coolant pump.
This event was previously documented in NRC Inspection Report 50-313/9103;'
50-368/9103. The actuation of the protective relay was caused by vibration from a degraded racking motor that was being used to rack up an adjacent breaker.
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The protective relay settings were ovaluated to determine the maximum
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allowable tolerances in ordcr to lower the sensitivity of the relay to
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vibration.
Procedure 2412.05, Revision 0, " Test and Inspection
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GE Models IAC66K, PJCl2D & West Model ITH Relays," was written, approved, and
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incorporated the necessary gap measurements.
The degraded racking. motor was
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removed from service and inspected. The cause of the vibration-from the
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racking motor was determined _ to be a worn collar located on the end of the i
motor shaft which engages the lifting mechanism for the breaker.
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Auxiliary operators involved in breaker operations were given additional training concerning breaker racking motor conditions and performance, and to ensure that all necessary precautions were fully uaderstood.
Based on the inspectors' review of Condition Repcrt 2-91-0060, Procedure 2412.055, and NRC Inspection Repo-t 50-313/91-03; 50-368/91-03, all corrective actions have been completed.
9.2 IClosed) Licensee Event Report 313/91-012:
" Automatic Actuation of the
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Control Room Emergency Ventilation S_ystem During Ventilation System
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Maintenance Caused by High Airborne Activity Which Resulted from an
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i Inadequate Prejob Evaluation" This licensee event report involved an actuation of the control room emergency
'7 ventilation system (CREVS) that occurred during the performance of maintenance to. change the filters on the Unit 2 Radwaste Area Exhaust Ventilation System.
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The CREVS actuation occurred when the count rate on the control room ventilation Radiation Monitor 2RE-8750-1 exceeded its alarm setpoint.
Condition Report C-91-0125 was initiated.
It was determined that the airborne
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activity in the radwaste area exhaust ventilation airhandler was most likely caused by back leakage through the air handler inlet damper of fission gases i
discharged to the system from the condenser vacuum system.
Fission gases in
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the condenser vacuum system would occur during leakage of the Unit 2 steam
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generator tubes.
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The root cause for the CREVS actuation was failure to identify potential input sources from the system prior to beginning work.
Personnel involved in the-planning and radiological safety evaluation process were not aware of the condenser vacuum system discharging to the radwaste area ventilation system.
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Procedure 2104.035, " Fuel Handling and Radwaste Area Ventilation," was revised
to ensure Unit 2 condenser vacuum pumps were aligned to draw from the
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separator tanks instead of from the atmosphere; and the Boron Management l
Holdup Tanks 2T-12A/B/C/D were secured during isolation of the Radwaste Area
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Ventilation Exhaust System Exhaust Fans 2VEF-BA and 2VEF-8B.
The inspectors reviewed applicable piping and instrument drawings and found j
that all exhaust systems which discharge to the radwaste area ventilation
system were evaluated for potential airborne activity during maintenance
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activities for the system.
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Personnel involved in performing radiological safety evaluations were
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counseled about the importance of obtaining assistance from operations and/or system engineering personnel, when performing radiological safety evaluations.
Based on a review of Condition Report C-91-0125, Procedure 2104.035, " Fuel i
i Handing and Radwaste Area Ventilation," and applicable piping and instrument drawings, the inspectors concluded that all corrective actions have been completed.
9.3 (Closed) Licensee Event Report 313/91-013
" Automatic Actuation of the Control Room Emergency Ventilation System Due to a Valid Radiation Signal
Which Resulted From Leakage From a Letdown Filter Drain Valve" l
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This licensee event report involved an actuation of the CREVS that was f
i initiated by Radiation Monitor 2RE-8750-1 which was located in the control room ventilation intake ductwork. The control room was lined up in the
recirculation mode with the supply and exhaust fans in service supplying ventilation to the computer room.
j Condition Report C-91-0126 was initiated and the root cause of the event was
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determined to be a small reactor coolant system leak from a letdown filter
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drain valve which was subsequently isolated. The CREVS equipment functioned
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as designed and the radiation levels in the control room returned to normal.
l The control room supply isolation damper was verified to be working properly.
l Actions to calculate and confirm no offsite release from the penetration room
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exhaust system discharge which draws from the area around the filter room where the leak was located were performed.
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The licensee concluded that since the circumstances of the event were unique
and there were no previous actuations of the CREVS from this source, no corrective actions associated with the orientation of the penetration room i
exhaust system discharge, or the control room outside air intake were planned.
l Based on review of Condition Report C-91-0126, the inspectors concurred with the licensee's conclusion.
9.4 (Closed) Licensee Event Report 368/91-018:
" Service Water Valve Stuck In
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Intermediate Position Causes Service Water Loop To Be Inoperable Resulting in Entry into Technical Specification 3.0.3"
This licensee event report involved the cooling tower make-up isolation valve
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from Loop 1 service water which failed in midposition. This event was
previously documented in NRC Inspection Report 50-313/91-33; 50-368/91-33.
The valve was a boundary between the seismic service water system and the nonseismic auxiliary cooling water system.
Loop 1 service water was declared
inoperable and Technical Specification 3.0.3 limiting condition for operation l
was entered.
Emergency Feedwater Pump 2P-7A was out of service for routine
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maintenance; therefore, since Loop 1 service water supplies cooling water to i
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the Emergency Feedwater Pump 2P-7B, both trains of emergency feedwater were I
declared inoperable.
Hydrogen Analyzer 2C-128B was also out of service for i
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Hydrogen Analyzer 2C-1288 and Emergency Feedwater Pump 2P-7A was returned to service.
The failed valve was manually placed in a shut position under operations administrative control. All Technical Specification limiting conditions for operation were exited within the allowable time limits.
Valve 2CV-1543-1 was disassembled, inspected, and repaired during Refueling Outage 2R9. A cut or gouge was found on the laminated seal ring,
and damage was also noted on the seat and disc of the valve. The licensee
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could not determine the root cause for the existence of the cut or gouge.
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Based on the inspectors' review of Unit 2 Technical Specifications, NRC Inspection Report 50-313/91-33; 50-368/91-33, and Condition
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Report 2-91-0620 and JO 00859654, all corrective actions have been completed.
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ATTACHMENT
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1 PERSONS CONTACTED
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1.1 Licensee Personnel q
C. Anderson, Unit 2 Operations Manager
D. Boyd, Licensing Specialist
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S. Cotton, Radiation Protection and Radwaste Manager
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R. Douet, Unit 1 Maintenance Manager
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N. Harris, Unit 2 Maintenance Manager
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R. King, Licensing Supervisor
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D. Moore, Health Physics Operations Superintendent
M. Sellman, Plant ' perations General Manager J
J. Vandergrift, Unit 1 Plant Manager 1.2 NRC Personnel
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S. Campbell, Resident Inspector
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A. Gaines, Acting Resident Inspector The personnel listed above attended the exit meeting.
In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
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2 EXIT MEETING An exit meeting was conducted on July 13, 1993.
During this meeting, the
inspectors reviewed the scope and findings of the inspection as detailed in
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this report. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.
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