IR 05000397/1988002
| ML17279B016 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/12/1988 |
| From: | Bosted C, Johnson P, Sorensen R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17279B014 | List: |
| References | |
| 50-397-88-02, 50-397-88-2, NUDOCS 8805030362 | |
| Download: ML17279B016 (17) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report No.:
50-397/88-02 Docket No.:
50-397 Licensee:
Washington Public Power Supply System P.
0.
Box 968 Richland, WA 99352 Facility Name:
Washington Nuclear Project No.
2 (WNP-2)
Inspection at:
WNP-2 Site near Richland, Washington Inspection Conducted:
Ja ua y 21 - March 10, 1988 Inspectors:
Approved by:
R.
C.
So
, Resident Inspector P.
H. )o nson, Chief Reacto& Projects Section
.. j C. J Boste'e or Resident Inspector I
I
'irJ'
I te igned Dat i ned dr Date Signed
~Summar:
Ins ection on Januar 21 - March 10 1988 50-397/88-02 Areas Ins ected:
Routine inspection by the resident inspectors of licensee action on previous inspection findings, control room operations, engineered safety feature (ESF) status, surveillance program, maintenance program, plant security, radiological protection practices, design review deficiencies, general employee and licensed operator simulator training, reactor building roof failure, licensee event reports, and special report reviews.
During this inspection, Inspection Procedures 30702, 30703, 37702, 61726, 62703, 71707, 71709, 71710, 71881, 90712, 90713, 92700, 92701, 92702 and 93702 were covered.
"
Results:,Two violations were identified, involving ineffective corrective actions in removing foam filters from the Class 1E switchgear cabinet (paragraph 3) and failure to perform an adequate design review of design change packages associated with Anticipated Transient Without Scram (ATWS)
modifications (paragraph 10).
Four previous followup items and three completed licensee event reports were closed.
8305030362 380418 PDR ADOCK 05000397 Q
DETAILS Persons Contacted Washin ton Public Power Su l
S stem
¹D. Mazur, Managing Director J.
Shannon, Deputy Managing Director
¹L. Oxsen, Assistant Managing Director for Operations
¹R. Glasscock, Director,- Licensing and Assurance
¹J.
Burn, Director, Engineering
~¹C.
Powers, Plant Manager
" J.
Baker, Assistant Plant Manager R. Corcoran, Operations Manager
" K.
Cowan, Technical Manager
" 0.
Feldman, Plant guality Assurance Manager R. Graybeal, Health Physics and Chemistry Manager J.
Landon, Maintenance Manager J.
Peters, Administrative Manager P.
Powell, Licensing Manager
- J.
Harmon, Assistant Maintenance Manager S.
McKay, Assistant Operations Manager L. Harrold, Generation Engineering Manager Nuclear Re ulator Commission
¹J.
Crews, Senior Reactor Engineer, Region V
~ Attended the Exit Meeting on March 10, 1988.
¹Attended the supplemental Exit Meeting on March 4, 1988.
The inspectors also interviewed various control room operators, shift supervisors and shift managers, engineering, quality assurance, and management personnel relative to activities in progress and records.
Plant Status At the start of the inspection period, the plant was operating at approximately lOOX power.
The plant operated at this power level until February 4.
An operator error during a surveillance test of the condenser vacuum instrumentation resulted in a test trip being initiated on the channel
"B" main steam isolation channel before the "A" channel was reset.
The second trip shut the main steam isolation valves and caused a reactor scram.
All systems performed as designed.
The plant was maintained in hot shutdown unti 1 restart on February 6.
Following restart, the plant operated until high reactor water conductivity forced a manual scram on February 13.
Subsequent investigation showed that a
condenser tube rupture had occurred when a baffle plate associated with a feedwater heater drain line broke and impacted the tube.
Tke plant was taken to cold shutdown for condenser tube troubleshooting and repairs.
On February 14, following breaker inspection mandated by NRC Bulletin
. 88-01, the reactor building air supply fan inadvertently started and
overpressurized the reactor building.
This is discussed further in paragraph 14.
The plant remained shutdown through March 6 while repairs were performed on the roof.
The plant was restarted on March 6 and power was increased to approximately 80K on March 8.
Power level increases were made within fuel preconditioning guidelines and, after control rods were adjusted for 100K operation, the power level was increased to lOOX by the end of the inspection period.
Previousl Identified NRC Ins ection Items 92701)
The inspectors reviewed records, interviewed personnel, and inspected plant conditions relative to licensee actions on previously identified inspection findings:
a.
Closed Enforcement Item 397/87-26-01:
Steam Tunnel Tem eratures Chan ed Without Prior NRC A
royal The setpoints for the steam tunnel high temperature trips were set higher than specified by the Technical Specifications without prior NRC approval.
The licensee obtained a revision to the Technical Specifications on December 15,1987, that allowed the trip setpoints to be raised.
Procedures which implemented the revision were issued in January 1988.
The inspector reviewed Plant Procedures Manual (PPM) 1.4.3,
"Revision of Master Data Sheets and Setpoints";
PPM 1.4.4, "Plant Instrumentation Design Documents";
PPM 1.4. 12, "Instrument Set Point Calibration".
These revisions appeared to implement the revised Technical Specification.
This item is closed.
b.
Closed Enforcement Item 397/87-26-02
Ke s to Hi h Radiation Areas Not Pro erl Controlled The inspector found the keys to the locked high and high-high radiation areas left unattended in the power plant.
Plant management issued a letter to the staff that discussed this item and.revised PPM 1.7. 1 "Access Key Control" and PPM 11.2.7.3
"Entry Into and Exit from High Radiation Areas".
The inspector reviewed the procedure revisions.
No additional key control problems have been experienced since this event.
This item is closed.
Closed Enforcement Item 397/87-19-19
Failure to Install Backu Nitro en Bottles in Accordance With Desi n Drawin s The backup nitrogen bottles were previously housed in a seismic structure, but the drawings required that the bottles be placed in the structures such that minimal movement was allowed.
The bottles were not the right size and were not shimmed so that the neck of each bottle was securely held by the rack.
The licensee obtained new MIL Specification bottles and installed these bottles in place of the identified bottles.
A procedure was
d.
installed on the wall near the bottles which describes how the bottles are to be changed.
The inspector examined these new bottles and read the posted instructions.
This item is closed.
(Closed Enforcement Item 397/87-19-31):
Foam Fi 1 ter s Instal 1 ed Without Pro er Modification Review Foam filters were found installed in the back of the residual heat removal (RHR) pump 1A 4160 VAC breaker on the outside surface.
These filters were not installed using a maintenance work'order, as requ'ired by station procedures, nor was the modification properly reviewed as required by 10 CFR 50'9.
This was identified to the license in Inspection Report 87-19.
On February 3, 1988, the inspectors observed that a foam filter was still installed on the inside of RHR-PlA br eaker cubicle and that foam filters were installed in the high pressure core spray (HPCS)
diesel generator exciter cabinet.
These filters had apparently been missed during the licensee's inspection of the switchgear.
The inspector brought this matter to the attention of plant management and both sets of filters were removed.
The inspectors subsequently verified that all filters had been removed.
This failure to take effective corrective action is a violation of 10 CFR 50 Appendix B.
This item is closed and further action will be taken under Enforcement item (88-02-01).
4.
0 erational Safet Ver ification 71707 a ~
Plant Tours The following plant areas were toured by the inspectors during the course of the inspection:
Reactor Building Control Room Diesel Generator Building Radwaste Building Service Water Buildings Technical Support.Center Turbine Generator Building Yard Area and Perimeter b.
The following items were observed during the tours:
(1)
0 eratin Lo s and Records.
Records were reviewed against Technical Specification and administrative control procedure requirements.
(2)
Monitorin Instrumentation.
Process instruments were observed
.for correlation between channels and for conformance with Technical Specification requirement o)
for conformance with 10 CFR 50.54. (k), Technical Specifications, and administrative procedures.
(4)
E ui ment Lineu s.
Valves and electrical breakers were verified to be in the position or condition required by Technical Specifications and Administrative procedures for the applicable plant mode.
This verification included routine control board indication reviews and conduct of partial system lineups.
During a tour of the control room on February 29, the inspector noted that an alarm on the station safety related 125 VDC battery ground was intermittently occurring.
This alarm was identified as being caused by personnel installing a modification for the "Anticipated Transient Without Scram" (ATWS) alternate rod insertion design change.
When questioned about why the ground was occurring, the system engineer told the inspector that he believed the ground was caused by connecting two safety related 125 VDC busses together.
This is discussed further in paragraph 10.
(5)
E ui ment Ta in
.
Selected equipment, for which tagging requests had been initiated, was observed to verify that tags were in place and the equipment was in the condition specified.
(6)
General Plant E ui ment Conditions.
Plant equipment was observed for indications of system leakage, improper lubrication, or other conditions that would prevent the system from fulfillingits functional requirements.
(7)
Fire Protection.
Fire fighting equipment and controls were observed for conformance with Technical Specifications and administrative procedures.
(8)
Plant Chemistr
.
Chemical analyses and trend results were reviewed for conformance with Technical Specifications and administrative control procedures.
(9)
~Securit
.
Activities observed are discussed in paragraph 8.
(10) Plant Housekee in
.
Plant conditions and material/equipment storage were observed to determine the general state of cleanliness and housekeeping.
Housekeeping in the radiologically controlled area was evaluated with respect to controlling the spread of surface and airborne contamination.
(11) Radiation Protection Controls.
Activities observed are discussed in paragraph 9.
On February 3, during a tour of the Control Room, several activated alarm indicators were discussed with the on-duty operators.
Alarms for the
"Rad Waste Building 432'orth Corridor" and "Reactor Building Exhaust Air Hi-Hi" alarms were activated, and operators initially told the inspector what they thought were the causes of the alarms.
Additional questioning by the inspector revealed that
the Rad Waste corridor alarm had occurred on the previous shift and the relieving shift had not,followed up on the alarm.
Operators then contacted the radiation protection department, the area was surveyed, and a line causing the radiation alarm was flushed.
This action caused the alarm to clear.
The reactor building exhaust air alarm was thought to be due to maintenance being performed on the strip chart recorder for that instrument.
The setpoint of the alarm was checked on the alarm response procedure and the control room log sheet.
A discrepancy associated with the alarm setpoint was noted between the two procedures, and members of the operations staff later determined that the alarm response procedure was in error.
The setpoint was corrected through a procedure change notice.
Operations and maintenance personnel later verified their initial belief that the alarm was caused by a missing jumper wire that activated the alarm when the strip chart recorder was removed.
Technicians replaced the jumper wire and returned the recorder to normal operation following completion of the maintenance.
The inspector observed on February 4 that the "Rad Waste 432'orth Corridor" annunciator was again activated.
The operators were aware of the condition causing the alarm, but had not initiated action to correct the condition (request radiation protection personnel to survey the area and flush the line again).
These items were discussed with plant management, and followup on these items was completed several days later.
Additional questioning of various control room alarms at several later times showed that the operators understood the causes of the alarms.
No violations of NRC requirements or deviations were identified..
5..
En ineered Safet Feature S stem Walkdown 71710 Selected engineered safety feature systems (and systems important to safety)
were walked down by the inspectors to confirm that the systems were aligned in accordance with plant procedures.
During the walkdown of the systems, items such as hangers, supports, electrical power supplies, cabinets,and cables were inspected to determine that they were operable and in a condition to perform their required functions.
The inspectors also verified that the system valves were in the required position and locked as appropriate.
The local and remote position indication and controls were also confirmed to be in the required position and operable.
Accessible portions of the following systems indicated date.
~Setem Diesel Generator Systems, Divisions 1, 2, and 3.
Hydrogen Recombiners Low Pressure Coolant Injection, (LPCI)
Trains "A", "B" and "C" were walked down on the Date January 31, February
February 2,
January 26, 29, February 3,
Low Pressure Core Spray (LPCS)
January 26, 29, February 3, 22, March 8 High Pressure Core Spray (HPCS)
Reactor Core Isolation Cooling (RCIC)
February
January 26, 29, March 8 Residual Heat Removal (RHR), Trains
"A" and "B" February 18,
Standby Liquid Control (SLC) System February 18,
Standby Service Water System February
125V DC Electrical Distribution, Divisions 1 and
February 3, March 7 250V DC Electrical Distribution February
No violations of NRC requirements or deviations were identified.
6.
Sur vei 1 lance Testin (61726 a ~
Surveillance tests required to be performed by the Technical Specifications (TS) were reviewed on a sampling basis to verify that:
1) the surveillance tests were correctly included on the facility schedule; 2)
a technically adequate procedure existed for performance of the surveillance tests; 3) the surveillance tests had been performed at the frequency specified in the TS; and 4) test results satisfied acceptance criteria or were properly dispositioned.
Portions of the following surveillances were observed by the inspectors on the dates shown:
Procedure 7. 4. 4. 1. 2 7. 4. 5. 1. 4
~ 7. 4. 3. 4. l. 1 Jet Pump Operability Check LPCI Valve Lineup Check ATWS Actuation Instrumentation 7;4.3.7. 12.34 Radiation Inst. Calibration Check Dates Performed January
January
January
February
7. 4. 6. 4. 1. 2 Suppression Chamber Vacuum Breaker Operability Test February
7.4.8. l. 1.2. 14 Diesel Generator 1,
18 month Inspection February
8. 3. 98 Functional Test of Reactor Building Ventilation Fans 7.4.8. 1. 1.2. 15 Diesel Generator 2,
18 month Inspection 7.4.8.2. 1.20 Weekly Battery Testing February
March 2,
March 3,
No violations of NRC requirements or deviations were identified.
7.
Plant Maintenance 62703 During the inspection period, the inspectors observed and reviewed documentation associated with maintenance and problem investigation activities to verify compliance with regulatory requirements, compliance with administrative and maintenance procedures, required QA/QC involvement, proper use of safety tags, proper equipment alignment and use of jumpers, personnel qualifications, and proper retesting.
The inspectors verified that reportabi lity for these activities was correct.
a.
The inspectors witnessed portions of the following maintenance activities:
Descri tion Dates Performed Calibration on MS 23D per AT 3122 Calibration on MS 23A per AT 3166 Remove/Rewor k REA-FN-1A per AV 1536 Hardware Installation in P650 Panel per AT 2677 January
January
February
February
Check/Verify REA "B" Loop Instrumentation per AT 3495
'February
Replace Suppression Pool Temperature Monitor February
per AT 1511 Removal of Reactor Building Roof per AT 3514 February 16-26 Repair Leaking Condenser Tubes per AT 3454 February 15-25 Installation of Supporting Structures for Reactor Building Roof Repair per AT 3499 February 22-26 Fabrication 8 Installation of Temporary Cover for Spent Fuel Pool per AT 3501 Febr uary 21-24 Installation of Structural Steel for Reactor February 21-26 Building Roof per AT 3502
Replacement of Piston Cooling Tube on Diesel Generator 2 per AT 3711 March
Rework REA-FN-1B per AT 2653 Bearing Replacement on REA-FN-lA per AT 2653 March
March 4 b.
While observing work associated with MWR AT 1511 (replacement of a channel selector switch), the inspector noticed that a
determination/retermination sign-off sheet was not being properly utilized.
The inspector observed that the initials for the retermination of a large number of leads were in place prior to the verifier's initials for the original determination.
Although the inspector could not verify the statement from available documentation, the assisting technician indicated that he had verified the determination as it was done and would place his initials on the check-off list when installation of the switch was complete.
While the method for documentation of independent verification was not specifically spelled out by procedure, plant management concurred that this practice was not meeting the intent of the procedure, nor of their expectations.
Plant management indicated to the inspector that action would be taken to ensure that these sign-off sheets are used as intended.
This issue will be followed up during the next inspection period as part of routine maintenance observation.
(88-02-02)
No violations or deviations were identified.
8.
Ph sical Securit (71881 The inspectors periodically observed security practices to ascertain that the licensee's implementation of the security plan was in accordance with site procedures.
The inspectors observed that the number of guards was adequate for the requirements of the security plan; that the search equipment at the access control points was operational; that the protected area barriers were well maintained without breaks; and that personnel allowed access to the protected area were badged and monitored and the monitoring equipment was functional.
Night illumination inside the protected area was observed and obstructions were lighted adequately.
Surveillance equipment was also observed during this inspection.
No violations of NRC requirements or deviations were identified.
9.
Radiolo ical Protection Practices 71709 The inspectors periodically observed radiological protection practices to determine whether the licensee's program was being implemented in conformance with facility policies and procedures and in compliance with regulatory requirements.
'reas observed included control point operation, and records of the licensee's surveys and postings of radiation, high radiation, and contamination areas within the radiological controlled area.
The inspectors also observed compliance with Radiation Exposure Permits, proper wearing of personnel monitoring
devices, and personnel frisking practices.
The inspectors verified that health physics supervisors and professionals conducted frequent plant tours to observe activities in progress and were generally aware of significant plant activities, particularly those related to radiological conditions and/or challenges.
ALARA (radiation exposures as low as reasonably achievable)
consideration was given to maintenance activities observed by the inspector.
No violations of NRC requirements or deviations were identified.
10.
Desi n Review Deficiencies While observing the implementation of a plant modification per ATWS design change package (DCP) 86-0229-0B, the inspector became aware of a ground alarm on the vital 125 VDC busses.
The engineer in charge of implementing the design change informed the inspector that it appeared that division 1 and division 2 were being cross-connected by the plant modification.
The engineer stopped work on the modification and began investigating the cause of the problem.
When the inspector followed up on the cross-connection of the two class 1E trains by the modification, the engineer informed him that several other discrepancies had also been noted.
In addition to the above actual cross-connection, the design had allowed other cross-connections between division 1 and 2 conductors.
The following additional design abnormalities were identified by the system engineer:
o In the event of an ATWS, the design called for the new vent and purge valves to open simultaneously, potentially negating the operation.
o Valve positions for the new valves would not indicate because the negative return was not included in the design.
o The design did not comply with the NRC's safety evaluation of the generic design
"NEDE 31096-P, Anticipated Transient Without Scram",
in that this document 'does not allow automatic reset after clearing of the actuation signal.
Although manual reset was the only method permitted, the designer had provided automatic reset only.
The inspector conducted followup inspection of samples of other design change packages that had been sent to the plant for implementation during the 1988 refueling outage.
In discussions with members of the plant technical staff who act as the modification project managers, the following design deficiencies were described to the inspector as noted:
o PMR 87-383-OA The FSAR includes commitments to have the reactor recirculation (RRC)
pump sensor and trip devices powered from different sources:
this was not implemented.
During the lOCFR 50.59 review, the reviewer checked off that the FSAR did not have any requirement o PMR 85-0718 100-pound air was to be supplied to purge the Radwaste building exhaust air sampler.
The sampler normally operates in a vacuum and has a
5 psi relief valve on the system.
As designed, the 100 psi air purge could potentially backfeed through a
common effluent header and overpressurize the Reactor Building effluent monitors.
o PMR 84-1505-2 An oil skid was added to the sample pumps for the continuous air monitors and sent to the plant without an electrical design package.
Both the pump motors and the logic switches to operate the system need electrical power to function.
The electrical portion of the design change package had been omitted during the design process.
The above discrepancies were identified by the licensee's plant staff during the design implementation phase, such that plant systems were not placed into service with the improper modifications installed.
In each case, the DCP had to be returned to Generation Engineering for additional engineering work.
The plant's design control process requires that the designer send the completed design package to an independent
"checker" for review before being sent to the cognizant engineer for design verification.
After design verification by the cognizant engineer, the design package is further reviewed by the Engineering Department group discipline manager and the plant operations committee (POC).
Following the POC review, the design package is issued to the plant staff for implementation.
The review by the plant technical staff is not a requirement of the design control process.
The design-review function of the design control process was inadequate in the above cases, in that documents with inappropriate designs were reviewed, approved, and issued to plant staff personnel for installation.
The inspector observed, however, that plant management had not taken an active role in resolving the issues or identifying the principal reasons for the repeated examples of poor engineering performance.
Licensee management was advised that failure to perform an adequate design review was considered a violation of the requirements of 10 CFR 50, Appendix B.
(Enforcement Item 88-02-03).
Corrective actions by members of the management staff was ongoing at the end of the inspection period.
The final corrective actions by the Engineering and the guality Assurance organizations will be evaluated during a future inspection.
11.
Post Tri Review/Root Cause Assessment A special supplemental inspection was accomplished that evaluated the post trip review and root cause assessment programs.
The inspector reviewed the most recent revision of the post trip review (PTR) procedure and the final draft of the root cause assessment procedure, and interviewed several members of the plant staff responsible for
implementation of the PTR process.
The inspector also noted that the PTR procedure was recently revised to include lessons learned, and several members of the plant staff had been introduced to root cause analysis through a series of human performance evaluation system (HPES) seminars.
Several completed PTRs were also reviewed by the inspector.
The inspector noted that the plant staff appeared to be improving in the implementation of the PTR process.
It was observed that the PTR procedure delineates what efforts are necessary for plant restart when the cause of the scram has not been fully understood.
The inspector noted, however, that the latest revision needed to be reviewed further by management to evaluate the need for a continuing assessment of probable causes when the plant has been restarted with the cause of the event sti 1.1 unknown.
For other than the PTR portion, the inspector concluded that the root cause program'as just beginning to be implemented.
The inspector reviewed the final version of the assessment of the roof fai lure discussed in paragraph 14, and concluded that it appeared to be sufficiently detailed to cover causes and precursors of the event.
The inspector concluded that the final draft of the root cause procedure needed to be revisited, and felt that the procedure was not a stand-alone document.
The inspector discussed with management the fact that the more successful root cause programs within the region rely on a dedicated and trained root cause group within the plant staff who either perform the root cause evaluations or review and concur with the line staff root cause evaluations.
The inspector stated that the plant would likely not be satisfied with the results of the root cause program if it were implemented as defined in the existing procedures.
Management agreed to further evaluate the concept of using a dedicated group to perform root cause evaluations.
No violations or deviations were identified.
General Em lo ee Trainin GET The inspector observed, and participated in, the licensee's training for new employees.
This training was approximately two days in duration and covered the areas of health physics, security, and quality assurance.
A textbook was issued to each student which covered these topics.
The inspector concluded that the lesson plan was organized, the instructor was prepared for administering the training, and the training was adequate.
In the area of quality assurance, considerable emphasis was placed on the individual's responsibility for assuring quality; however, very little emphasis was placed on how an individual's potential quality concerns were to be addressed.
Little discussion was presented concerning how a nonconformance report (NCR) was to be initiated, or how the licensee's Direct Line was to be used.
This apparent weakness was discussed with the Training Manager.
He indicated that a copy of an NCR would be included in the course textbook and that the course would also be changed to familiarize students with its initiation and use.
The inspector also
noted that the phone number for the Direct Line was not included in the textbook.
The Training Manager stated that this would be considered.
Finally, the inspector later learned that some guidance given in the GET concerning dosimetry was incorrect.
The actual plant practice was to sign in on a dose card whenever an RWP was signed, although Training had indicated that a dose card was to be signed only when a radiation area was physically entered.
Plant HP representatives stated that action had been taken to correct the GET in this area.
No violations or deviations were identified.
13.
Observation of Licensed 0 erator Trainin on the Plant Simulator The inspector observed portions of the licensee's hands-on training for licensed operators that was conducted on the plant simulator.
This phase of training was somewhat informal with the instructors actively participating.
It was composed of various hypothetical scenarios that depicted abnormal operational situations.
One scenario involved loss of power to bus SM-2 during a reactor startup with a subsequent loss of feedwater.
The other scenario involved a severe steam leak in containment.
The operators responded well to both of these events.
Short critiques were conducted at the end of each scenario which evaluated the operators'esponses.
In the case of the steam leak, the control room supervisor did not use the flow chart for accident situations due to its distant location from the control room panels.
The inspector mentioned this observation to the Plant Manager and questioned whether the operators understood what was expected of them -- specifically, whether they were expected to use the flow charts.
Plant management stated that existing policies permitted the operators to use either the flow chart or the approved procedures, but that this would be evaluated further.
No violations or deviations were identified.
14.
Reactor Bui ldin Roof Failure 93702 On February 14, while in a cold shutdown condition, technicians conducted an inspection of certain electrical breakers in accordance with NRC bulletin 88-01.
Operators were in the process of restoring electrical breakers to service for the reactor building supply and exhaust air fans when the reactor building air supply fan ROA FN 1A inadvertently started without the trip circuit being energized.
All protective features and fan status indicating lights are powered from the breaker's trip circuit power supply.
(When the equipment operator restored the breaker to operating status, he did not notice that the indicating lights were not energized.
The operating instruction was later reviewed and it was determined that this procedure did not specify that the lights were to be checked.
This condition was noted by plant management and a change to the instruction was initiated.)
The fan apparently started when the operator moved the fan breaker control switch for ROA FN 1A from the pull-to-lock position.
The control room operator noted that the breaker
"open" light for the ROA fan was not lighted, and assumed that the light bulb was burned out.
He then returned the breaker control switch to the pull-to-lock position and attempted to replace the "open" light bulb.
The operator did not have indication that the breaker was actually closed, and no further abnormal conditions were evident unti 1 the control room supervisor (CRS) noted that the secondary containment pressure recorder had indicated off-scale (full-scale was 7 in.
H 0) and subsequently decreased to 1.5 in.
After attempts to adjust the )ressure were unsuccessful, an equipment operator (EO) was sent to check fan ROA FN 1A, which was determined to be running.
The EO then manually tripped the fan breaker at the direction of the control room operator.
The cause of the fan starting was later determined to have been miswiring of the control circuit while installing a plant modification during plant construction.
Previous inadvertent starts had been observed by plant operations personnel, and a work request had been submitted in 1985, but troubleshooting of the problem had been inconclusive.
The operating procedure for starting the ROA/REA (reactor building supply/exhaust)
fans was modified at that time, however, to warn the operator that the fans may autostart when the control switch is taken out of the pull-to-lock position.
At approximately 11:30 p.m.
on February 14, during his initial tour of the plant after coming on shift, an equipment operator noticed that several nuts and bolts were lying on the refueling floor.
Subsequent investigation revealed that the secondary containment roof was damaged.
This roof covers the equipment housed on or above the reactor refueling floor including the spent fuel pool.
The damage assessment performed by plant staff on the morning of February 15 indicated that most of the roof panels had moved, some as much as 9-12 inches, and that several panels had not fully reseated.
Two panels were displaced approximately eight inches above their normal height.
These panels were designed to relieve in the event of excessive internal pressure, with the shear bolts which secured the panels designed to break at shear forces equivalent to 14 in.
(H 0) building pressure.
Based on an engineering review of'available event data, the licensee concluded that the maximum building pressure reached during the event was approximately 13.5 in.
The licensee performed a preliminary inspection of the spent fuel pool, which was located under one of the displaced panels, and determined that some debris had fallen into the spent fuel pool.
The licensee did not have any indication that any of the spent fuel stored in the pool was damaged.
During the first week of the roof repair effort, a temporary'over was erected over the spent fuel pool to prevent further debris from falling into the pool.
During the second week of repairs, a Technical Specification change was obtained that allowed a temporary missile shield to be placed over the spent fuel pool to prevent any heavy objects from falling into the spent fuel pool during roof repair The licensee's staff performed additional inspections of other plant equipment, including primary containment penetrations, plant instrumentation, and secondary contai'nment vertical walls for potential adverse effects of building overpressurization.
The staff did not identify any additional damage to any other systems.
Repairs were completed to the roof and secondary containment was reestablished on March 2.
The missile shield covering the spent fuel pool was removed on March 3, and preparations were made to proceed with plant restart after completion of functional tests for the fan controls and protective devices.
Upon functionally testing the exhaust and supply fans on March 3, it was discovered that the exhaust fan (REA FN 1B) would trip approximately 1-2 minutes after starting and that the annunciator and computer alarms did not function properly on an automatic start of the standby fan.
At this time, the REA FN lA fan bearings were being replaced by the maintenance department.
Tests on the 1B breaker, which had been replaced from spares (as a result of inspections performed pursuant to NRC Bulletin 88-01),
indicated that the thermal over load devices were set conservatively.
Tripping of the 1B REA fan was determined to be a function of this lower setpoint, but primarily a result of the test conditions (secondary containment doors open to atmosphere),
and not of concern during normal system operation.
The lack of annunciator and computer alarms was determined to be due to the short time duration between the sensing of a condition causing a fan trip and the time when the relay scheme tripped the malfunctioning fan, isolating the problem.
The licensee determined that a signal duration of approximately 40 milliseconds was necessary to permit the annunciators to function.
The thermal overload device for the fan 1B breaker was reset to a higher setpoint and the alarm indicator relay for the computer and alarm annunciator was modified to provide a signal duration of approximately
seconds, which allowed the alarms to register.
One remote annunciator problem was also traced to a missing jumper on an annunciator circuit card, which was corrected by the licensee.
.A successful functional test was completed on March 5 and the-Reactor Building ventilation system was declared to be operational.
The inspector witnessed both the unsuccessful functional test on March 3 and the satisfactory test on March 5.
Additional details of this event and the licensee's root cause assessment were.as described by the licensee in LER 88-07.
This event and the licensee's corrective actions were also discussed, prior to plant restart, in a March 1, 1988 management meeting.
During this meeting, as documented in meeting report no. 50-397/88-07, NRC representatives expressed concern that plant staff personnel proceeded with plant evolutions before anomalous equipment behavior or conditions were understood.
Licensee management agreed to reemphasize this concern prior to plant restart.
Other ongoing licensee activities will be examined during followup review of the LER.
No violations or deviations were identifie.
Licensee Event Re ort (LER) Followu 90712 72700 The following LERs associated with operating events were reviewed by the inspectors.
Based on the information provided in the reports it was concluded that reporting requirements had been met, root causes had been identified, and corrective actions were appropriate.
The following LERs are closed.
LER 87-15 Mode change while Division 2 Drywell Pressure monitoring instrumentation was inoperable.
LER 87-16 Spurious Actuation of Control Room Emergency Filtration System LER 87-21 Inadvertent Nuclear Steam Supply System (NSSS)
isolation during surveillance test.
No violations or deviations were identified.
16.
Review of Periodic and S ecial Re orts 90713 Periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9. 1 and 6.9.2 were reviewed by the inspector.
This review included the following considerations:
the report contained the information required to be reported by NRC requirements; test results and/or supporting information were consistent with design predictions and performance specifications; and the reported information appeared valid.
Within the scope of the above, the following reports were'eviewed by the inspectors.
o Monthly Operating Report for January, 1988.
o Monthly Operating Report for February, 1988.
o Semi-Annual Effluent Report for July 1 - December 31, 1987 No violations or deviations were identified.
17.
Exit Meetin 30703 The inspectors met with licensee management representatives periodically during the report period to discuss inspection status, and an exit meeting was conducted with the indicated personnel (paragraph 1) on March 10, 1988.
A supplemental exit meeting, with Mr. J.
Crews of Region V
participating, was held on March 4 to discuss the root cause, post trip review, and design review problems identified above.
The scope of the inspection and the inspector's findings, as noted in this report, were discussed and acknowledged by the licensee representatives.