IR 05000528/1990003
| ML17305A712 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 04/12/1990 |
| From: | Wong H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17305A710 | List: |
| References | |
| 50-528-90-03, 50-528-90-3, 50-529-90-03, 50-529-90-3, 50-530-90-03, 50-530-90-3, NUDOCS 9004270015 | |
| Download: ML17305A712 (65) | |
Text
U ~
S.
NUCLEAR REGULATORY COMMISSION
REGION V
Re ort Nos.
Docket Nos.
License Nos.
Licensee:
Faci1 it Name:
50-528/90-03, 50-529/90-03 and 50-530/90-03 50-528, 50-529, 50-530 NPF-41, NPF-51, NPF-74 Arizona Public Service Company P.
0.
Box 52034 Phoenix, A2 85072-2034 Palo Verde Nuclear Generating Station Units1, 2&3 Ins ection Conducted:
January 21 through March 3, 1990 Inspectors:
Approved By:
DE Coe, R. Barr, T. D'Angelo, C. Myers, F.
Ringwald, J.
Sloan, Senior Resident I'nspector Senior Resident Inspector (Trojan)
Senior Resident Inspector (Rancho Seco)
Resident Inspector (Rancho Seco)
Resident Inspector Resi ent Inspector p'> go ong, Reactor Projects Branch,Section II Ins ection Summar
te igne Ins ection on Januar 21 throu h March
1990 (Re ort Numbers an
Areas Ins ected:
Routine, onsite, regular and backshift inspection by the t ree ress ent inspectors, plus various inspectors from the Region V staff.
Areas inspected included: previously identified items; review of. plant activities; engineered safety features system walkdowns; monthly surveillance testing; monthly plant maintenance; core protection calculator (CPC)
surveillance test plug - Unit 1; containment hydrogen monitor surveillance-Unit 1; nuclear construction department (NCD) performance of work order steps
- Unit 1; improper fire watch performance - Unit 1; load shed potential transformer failure - Unit 1; fuel building rollup door damage/ventilation damper jumper installation - Unit 1; fuel building radiation monitor modification engineered safety features (ESF) actuation - Unit 1; low pressure safety injection (LPSI) "B" pump high differential pressure (D/P) - Unit 1; loss of instrument air - Unit 1; heated junction thermocouple (HJTC) removal-Unit 1; diesel generator fuel oil storage tank level meter inaccuracy-Unit 2; preparation for refueling - Unit 2; inappropriate
CFR 50.59 review
- Unit 3;
CFR Part 21 Program - Units 1, 2, and 3; allegation followup-(ATS No. RV-89-A-0030); allegation followup - (ATS No.
RV-89-A-'0067);
r
allegation followup -
(ATS No.
RV-89-A-0068); allegation followup - (ATS No.
RV-89-A-0069); review of licensee event reports Units 1, 2 and 3; and review of periodic and special reports - Units 1, 2 and 3.
During this inspection the following Inspection Procedures were utilized:
30702, 30703, 36100, 61726, 62703, 71707, 71710, 90712, 90713, 92700, 92701, 92702, and 93702.
Results:,
Of the 25 areas inspected, one violation was identified.
The
~v>o at>on pertained to the improper uee of a gage at Unit 2.
General Conclusions and S ecific Findin s
Si nificant Safet chatters:
Summar of Violations:
Summar of Deviations:
0 en Items Summar
1 Violation 7 items closed, 1 item left open, and 4 new items opene DETAILS Persons Contacted The below listed technical and supervisory personnel were among those contacted:
Arizona Public Service Com an (APS)
R. Adney,
"J. Allen,
- J. Bailey, B. Ballard, F.
Buckingham, H. Bieling,
- T. Bradish, P. Brandjes,
~P. Caudill,
"T. Cogburn, W.
Conway, D. Fasnacht, E. Firth,
- D. Heinicke, P;
Hughes,
"W. Ide, S.
Johnson, F. Larkin,
"J.
Levine, J.
LoCicero,
"W. Harsh, D. Hauldin, C.
Rogers, C.
Russo, G. Shell, W. Simko,
- E. Simpson,
"G. Sowers, D.
Swan,
"S. Terrigino, P. Wiley, R.
Younger, The inspectors al during the course Plant Manager, Unit 3 Engineering 8 Construction Director Vice President, Nuclear Safety 8 Licensing equality Assurance Director Operations Manager, Unit 2 Emergency Plan/Fire Protection Manager Compliance Manager Central Maintenance Manager Site Services Director Standards and Tech.
Support Director Executive Vice President - Nuclear Nuclear Construction Manager Training Manager Plant Manager, Unit 2 Radiation Protection 8 Chemistry Manager Plant Manager, Unit 1 Participants Site Representative Security Manager Vice President, Nuclear Power Production Independent Safety Engineering Manager Plant Director Outage Planning 8 Management Manager Licensing Manager equality Control Manager equality Systems Manager Maintenance Manager, Unit 2 Vice President of Engineering 8 Construction Engineering Evaluations Manager Shift Supervisor, Unit 2 Management Services Supervisor Work Control Manager, Unit 2 Plant Standards and Control Manager so talked with other licensee and contractor personnel of the inspection.
"Attended the Exit meeting held with NRC Resident Inspectors on March 1, 199.
Previousl Identified Items - Units 1
and 3 (92701 92702 an a.
(Closed) Follow-u Item (528/529/530/88-18-P)
"Part 21 rac in o
i es in nc or ar in a ves
-
ni s
an b.
This Part 21 report concerned a problem of cracked slides in four-way valves furnished on the actuators of the main steam isolation valves.
This item was initially reviewed in Inspection Report 89-36 and left open pending review of the licensee's evaluation of the problem at Palo Verde.
During this inspection, the inspector reviewed Engineering Evaluation Request (EER) 88-SG-123 and found that the licensee had determined that inspection of the subject valves was required.
The licensee completed the inspection of all twelve valves (four per unit) and found one cracked slide which was subsequently replaced.
No other followup actions are planned by the licensee.
This item is closed.
Closed)
Follow-u Item (528//529/530/89-16-P)
"Part 21 Re ort of Limitor ue o
e
-
-
nits
2 an 3.
The Washington Public Power Supply System (Supply System)
reported a
defect with a Limitorque Model SMB-2 valve operator.
The report identified a casting defect in the upper housing cover which interfered with the rotating bearing edge of the upper thrust bearing.
The Supply System Part 21 report stated that Limitorque SMB-2 covers were not used on any other SMB model valve operators.
Based on a determination that no SMB-2 operators were installed in safety-related applications, APS determined no further evaluation was necessary.
In addition, the licensee indicated that this defect would show abnormalities in the diagnostic test signature (MOVATS)
related to thrust versus motor current.
The licensee's Compliance Manager, who is responsible for coordinating all 10 CFR Part
reports, stated that no such defects have been identified in'll MOVATs testing performed to date at Palo Verde.. This item is closed.
C.
(0 en) Enforcement Item (528/89-16-03)
"Ino erable Atmos heric um a ve o.
-
nit Y
The remaining actions committed to by APS are to upgrade operating procedures and train licensed and non-licensed operators o'
Procedure 02AC-OZZOl, "Independent Verification of Valves, Breakers and Components."
The operating procedure upgrade program was originally scheduled to be completed by December 1989.
The schedule as of December 1989, is:
"(1) A plan for the review, revision, and issuance of the Nuclear Administrative and Technical Manual Procedures
- Preliminary due date:
6/31/90 [sic].
(2)
Initiate revision of the affected Operations procedures-Estimated star t date: 8/1/90."
The inspector considered that this apparently shifts the due date for this project's schedule to June 1990, and does not contain a
completion date.
All currently licensed operators were trained as scheduled by October 30, 1989.
All current non-licensed operators were briefed by October 30, 1989-, but did not receive training until January 1990.
The incorporation of training on Procedure 02AC-OZZOl has not yet been incorporated into training materials for initial licensed or non-licensed training.
The training plans will be revised to meet this.commitment before they are used in training the next classes of operators.
This item will remain open unti 1 the procedure upgrade schedule is reviewed.
(Closed) Follow-u Item (528/89-36-02
"Ca acitor Bank 1 re nl The licensee revised their program for inventory of Aqueous Film Forming Foam and inspection requirements for auxiliary equipment on the fire truck.
The inspector reviewed these program revisions and inspected the foam storage and had no further questions.
This item is closed.
(Closed)
Unresolved Item (528/89-43-01
"Bent Airline on t e ent ue oo ate ea u
ines
-
nit 1.
The inspector reviewed the licensee's Incident Investigation Report 3-1-89-106, dated December 12, 1989, which reviewed the circumstances surrounding the bending of gate seal airlines, possibly by the spent fuel pool crane.
The licensee's investigators interviewed the people involved, and reviewed gC Zone III logs, work orders, procedures, the Work Control (SIMS) database and Engineering documents.
The licensee was not able to establish a root cause.
The licensee repaired the damage.
The consequences of the event were negligible and apparently isolated.
This item is closed.
(Closed Follow-u Item (530/89-16-02
"Reassessment of oo erations
-
nit This item resulted from several observed inadequacies in the licensee's conduct of midloop operations.
The inspector reviewed applicable licensee procedures and noted that they had been revised to incorporate lessons learned.
Additionally, the licensee documented that a review of these procedures relative to NRC Generic Letter 88-17 had been completed and 'all requirements reverified as being met.
Finally, the licensee documented in a letter to the NRC dated November 14, 1989 that testing of the Reactor Coolant and Shutdown Cooling Systems flows had been completed in Units 1 and
and would be performed during the Unit 2 refueling outage.
The
e analysis of these results would be used as necessary to further ensure that procedural guidance would be effective in preventing vortexing during midloop operations.
These actions appeared appropriate.
This item is closed.
(Closed)
Enforcement Item (530/89-16-03)
"Vortexin and
>r ntrasnment urban s
oo erations nit This item resulted when vortexing and air entrainment in the operating shutdown cooling loop occur red during midloop operations due to excessive shutdown cooling flow for the existing Reactor Coolant System (RCS) level.
The inspector reviewed the licensee's reply to the Notice of Violation and determined that procedure changes involving limiting the number of operating shutdown cooling loops, the maximum flow, and the performance of periodic venting appeared adequate to prevent recurrence.
Although the licensee further committed to additional Auxiliary Operator (AO) training, the inspector noted that midloop operation is not being planned by the licensee.
Should subsequent midloop operations be scheduled, operator refresher briefings would be required per the licensee's procedures.
This item is closed.
No violations of NRC regulations or deviations were identified.
3.
Review of Plant Activities (71707 and 93702)
a.
Unit 1 b.
Unit 1 remained in Mode 5 throughout this inspection period.
Reactor vessel head stacking continued but was not complete at the end of the inspection period.
The reactor coolant system (RCS)
was filled and vented, a bubble was drawn in the pressurizer and the reactor coolant pumps were run successfully.
The RCS was subsequently cooled back down to ambient temperature and depressurized for the Containment Integrated Leak Rate Test (ILRT),
which was completed satisfactorily.
Unit 2 Unit 2 entered this reporting period in Mode 1 at 100% and remained at essentially 100% power until February 22, 1990, when power level was lowered to approximately 90% due to an extended Core Operating Limits Supervisory System (COLSS) outage.
The plant reduced power on February 23 in preparation for its scheduled refueling outage, but the reactor was manually tripped from 25% at ll:01 PM, due to a large negative Axial Shape Index (ASI).
After the uncomplicated trip, the plant was cooled down, entering Mode 5 at 5:57 AM, on February 25, 1990.
The plant remained in Mode 5 through the end of the reporting period and commenced its second refueling outage.
Unit 3 Unit 3 began the reporting period in Mode 1 at approximately 18%
power performing various power ascension testing subsequent to their
first refueling outage.
At the end of the reporting period, Unit 3 had completed all required power ascension physics testing and was holding power at 98X or less pending an assessment of an apparent slight increase in total electrical megawatts for the same Cycle
thermal reactor power.
The licensee had not completed this evaluation by the end of the inspection period.
d.
Plant Tours The following plant areas at Units 1, 2 and 3 were toured by the.
inspector during the inspection:
o Auxiliary Building o
Containment Bui 1 ding o
Control Complex Building o
Diesel Generator Building o
Radwaste Building o
Technical'Support Center o
Turbine Building o
Yard Area and Perimeter The following areas were observed during the tours:
l.
0 eratin Lo s and Records - Records were reviewed against ec naca pec>>cat>on and administrative control procedure requirements.
In three instances in Unit 2, Control Room logs were found not to contain information regarding important plant evolutions.
These instances involved the venting of the pressurizer to the containment, the reinstallation of the containment equipment hatch prior to the pressurizer venting evolution, and the removal from service of a charging pump on February 14, 1990.
In these cases, Unit 2 management was informed of the inspector's concerns and acknowledged the need for better log keeping.
2.
Nonitorin Instrumentation
- Process instruments were observed
" for corre ation etween c annels and for conformance with Technical Specification requirements.
Shift Mannin
- Control room and shift manning were observed or con ormance with 10 CFR 50.54.(k), Technical Specifications, and administrative procedures.
E ui ment Lineu s - Various valves and electrical breakers were ver>
)e to e )n the position or condition'equired by Technical Specifications and administrative procedures for the applicable plant mode.
5.
E ui ment Ta in
- Selected equipment, for which tagging requests a
een initiated, was observed to verify that tags were in place and the equipment was in the condition specifie.
General Plant E ui ment Conditions - Plant equipment was o serve or sn >cat)ons o
system leakage, improper lubrication, or other conditions that would prevent the systems from fulfilling their functional requirements.
7.
Fire Protection Fire fighting equipment and controls were f
ithT hi 1Sp io ti d
administrative procedures.
8.
Plant Chemistr
- Chemical analysis results were reviewed for con ormance w>th Technical Specifications and administrative control procedures.
9.
~Securit
- Activities observed for conformance with regulatory requirements, implementation of the site security plan, and administrative procedures included vehicle and personnel access, and protected and vital area integrity.
The SAS was included during plant tours.
10.
Plant Housekee in
- Plant conditions and material/equipment s orage wer e o served to determine the general state of cleanliness and housekeeping.
11.
Radiation Protection Controls - Areas observed included control po)nt operation, recor s
o icensee's surveys within the radiological controlled areas, posting of radiation and high radiation areas, compliance with Radiation Exposure Permits, personnel monitoring devices being properly worn, and personnel frisking practices.
No violations of NRC requirements or deviations were identified.
4.
En ineered Safet Features S stem Walkdowns - Units
2 and 3 (71710 Selected engineered safety features systems (and systems important to safety)
were walked down by the inspector to confirm that the systems were aligned in accordance with plant procedures.
During this inspection period the inspectors walked down accessible portions of the following systems.
Unit 1 o
High Pressure Safety Injection (HPSI)
pumps o
Reactor Mater Tank o
"A" Emergency Diesel Generator Unit 2 o
"A" Emergency Diesel Generator Unit 3 o
"A" Emergency Diesel Generator
f
In Units 1 and 2, a chain fall was observed hanging from an overhead trolley near the generator end of the "A" Emergency Diesel Generator.
The chain was unsecured and could swing into contact with the 'generator casing.
In Unit 3, the chain was securely stored in a container at the trolley.
The Unit 1 and 2 Shift Supervisors initiated action to store the chain falls securely.
In Unit 2, a Diesel Generator Fuel Oil Storage Tank levelmeter was observed to be reading below the Technical Specification minimum of 80K due to apparent static electricity effects.
This is addressed in Paragraph 17 of this report.
No violations of NRC requirements or deviations were identified.
5.
Monthl Surveillance Testin
- Units 1 2 and 3 (61726 Selected surveillance tests required to be performed by the Technical Specifications (T/S) 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 T/S; and 4)
test results satisfied acceptance criteria or were properly dispositioned.
b.
Specifically, portions of the following survei llances were observed by the'inspector during this inspection period:
Unit 1 Procedure o 41ST-1RC01 Reactor Coolant System (RCS) Pressurizer Keatup and Cooldown Rates o 36ST-1SE01 Log Power Functional Test o 36ST-9HP04 Containment Hydrogen Monitor System Calibration Test o 41ST-1DG02 Diesel Generator
"B" Test 4.8. 1. 1.2.(a)
o 77ST-9SB04 Core Protection Calculator (CPC) Channel
"D" Calibration Unit 2 Procedure Descri tion o 32ST-9ZZ34 Battery Charger Surveillance Test o 73ST-9ZZ22 Snubber Test (2SI-072-H-016)
Unit 3 Descri tion 6.
Monthl Plant Maintenance - Units 1 2 and 3 (62703)
P d
o 43ST-3SG04 Testing Atmospheric Dump Valves in Mode 1 t.
No violations of NRC requirements or deviations were identified.
a.
During the inspection period, the inspector observed and reviewed selected documentation associated with maintenance and problem investigation activities listed below to verify compliance with regulatory requirements, compliance with administrative and maintenance procedures, required equality Assurance/equality Control involvement, proper use of safety tags, proper equipment alignment and use of jumpers, personnel qualifications, and proper retesting.
The inspector verified that reportabi lity for these activities was correct.
b.
Specifically, the inspector witnessed portions of the following maintenance activities:
Unit 1
Descri tion o
Refurbish Fire Door J-123 o
Installation of "B" Spray Pond Filter Pump Suction Breaker o
Replace Louver on Door G-102 between
"A" Diesel and Diesel Equipment Room o
Replace
"B" Diesel Generator Lube Oil Cooler Plug o
Troubleshoot
"B" Diesel Generator Lube Oil Temperature Switch o
Weld Repair Fire Door in "B" Switchgear Room to HVAC Chase Unit 2 Descri tion o
Troub1 eshoot RU-146 o
Troubleshoot Reactor Trip Breaker "8" o
Troubleshoot/Repair Lock on Control Room Door o
Troubleshoot/Repair Main Steam 5-Ton Monorail Hoist Controls
Unit 3 Descri tion o
Calibrate RU-153 No violations of NRC requirements or deviations were identified.
Core Protection Calculator (CPC Surveillance Test Plu - Unit 1 61726)
The.inspector observed the, performance of procedure 77ST-9SB04,
"Core Protection Calculator (CPC) Channel 'D'alibration." 'uring the calibration, the Isolator Test Plug could not be found in the box of plugs and cards the technicians brought to support this test.
The technicians expressed some confusion as to which plug might be the Isolator Test Plug and whether or not a particular plug was defective.
After telephoning the system engineer, the technicians obtained a plug which was not clearly labeled but which the technicians stated to be the
'correct test plug.
The surveillance procedure depicts the use of this plug without identifying the input pins, internal circuitry or output leads.
The inspector questioned whether special test equipment should be clearly and unambiguously labeled so there is no question that the correct device is being used, particularly when there are multiple different devices which fit interchangeably.
The inspector also questioned whether technicians should to be sure that special test equipment is functioning properly.
Licensee management committed that all test plugs will be clearly and prominently labeled.
Subsequent to the inspector's questions, the licensee determined that a defective Isolator Test Plug could inappropriately give satisfactory test results making the results indeterminate.
The licensee agreed to revise the surveillance procedure to include a step to verify the proper functioning of the Isolator Test Plug before performing the subsequent steps in the surveillance procedure.
No violations of NRC regulations or deviations were identified.
Containment H dro en Monitor Surveillance - Unit 1 (61726)
During the performance of Surveillance Test 36ST-9SB04, the technician received an analyzer cell failure alarm when the zero potentiometer was touched.
The zero potentiometer shaft was not rotated but merely pushed axially.
Subsequent to the inspector's questions, a work request was issued to correct this situation; however the alarm cleared on its own and the surveillance was completed and signed off as satisfactory with the work request still outstanding.
The inspector questioned whether the abnormal operation of safety equipment, such as unexpected alarms, should be fully evaluated and understood prior to signing off surveillances on that equipment and declaring it operable.
The licensee stated that all testing requirements of the surveillance test were met and the surveillance test had been signed off as complete even with the apparent short.
The inspector considers this an unresolved issue (Unresolved Item 528/90-03-01).
The inspector also noted that a valve hand wheel was lying on the floor of the hydrogen monitor.
The technician indicated that it had been unattached for at least six months prior to the Unit 1 shutdown one year ago.
The technician also indicated that the unattached hand wheel was not a problem because it could be picked up and used to manipulate the valve as needed.
Subsequent to the inspector's questions, the technician submitted a work request to attach the hand wheel.
The inspector further questioned whether plant workers should identify plant problems and take appropriate actions to assure the correction or resolution of the problem.
The licensee agreed that the ILC technician was not attentive enough to the broader perspective beyond the immediate task at hand.
The I8C supervisor discussed this with the technician and will be issuing a
memo to emphasize the need for all technicians to be alert for problems they encounter in the field and to input the problems into the work control program.
No violations of NRC regulations or deviations were identified.
Nuclear Construction De artment (NCD) Performance of Mork Order Ste s -
nit 1
3 Mhile installing siphon breakers in the spray pond filter piping, the inspector noted that the craft workers understood that they could complete several work order steps before these steps were signed off by the Construction Responsible Engineer (CRE).
The requirement for performing NCD work is in Procedure 82DP-OMPOl, Paragraph 3.8.28, which states "... the work shall not proceed to the next sequenced work step until documentation of the completed work step is made."
The licensee stated that when the order of work step performance is important, the work order explicitly states to perform the specified steps in the order listed.
These
'-'special category 'ork steps are considered
"sequenced" work steps and that Paragraph 3.8.28 only applies to these
"sequenced" steps.
The inspector noted that without a definition of "sequenced" steps the procedure appears to apply to all work order steps.
Licensee management agreed that this interpretation is not readily apparent and that the procedure would be revised to clarify times when strict work step completion must be followed.
No violations of NRC regulations or deviations were identified.
Im ro er Fire Match Performance
- Unit 1 (62703)
The inspector observed a weld repair to door J-115 in the "B" Essential Switchgear Room.
The inspector noted that the fire watch's fire extinguisher was approximately ten feet away from the fire watch and behind a pillar and a cart.
Mhen the inspector discussed this situation with the fire watch, the fire watch admitted that the fire extinguisher was too heavy to be lifted.
The inspector discussed this with the Nuclear Construction Department Manager and the Fire Protection Supervisor.
Subsequently, the licensee stated that this fire watch has been removed from fire watch duties.
The licensee has incorporated a
physical demonstration of a fire watch's ability to handle and use a fire
extinguisher during fire watch training and has written an instruction change request to include a verification of a fire watch's ability to handle and use an extinguisher in 14AC-OFP04, "Fire Watch Duties.
'o violations of NRC,regulations or deviations were identified.
Load Shed Potential Transformer Failure - Unit 1 (93702).
A load shed of 13.8KV bus NAN-S02 occurred in Mode 5 due to a failure of the Potential Transformer (PT) across the "B" to "C" phase of the bus, Three months earlier Unit 1'had experienced a fai lure of an identical potential transformer on NAN-SOl, but did not experience a load shed because the PT which failed was across the "A" to "B" phase of the 13.8KV bus and provides for bus synchronizing rather than a bus load shed function.
The licensee has written Engineering Evaluation Requests (EER) 89-NA-049 and 90-NA-002 to investigate the root cause of these failures.
The inspector concluded that these EERs -need to be complete before this issue can be completely evaluated.
This is a follow-up item (528/90-03-02).
As of the end of this inspection period, these EERs have not been completed, No violations of NRC regulations or deviations were identified.
Fuel Bui ldin Rollu Door Dama e/Ventilation Dam er Jum er Installation nest
While preparing for an upcoming PBA-S03 electrical outage, a work request was generated to provide air hose jumpers around actuators for dampers HFA-M01, HFA-M02, HFA-M03 and HFA-M04 in the Fuel Building ventilation system.
This work request was written to remove the air supply to a damper actuator for a short time, which would cause the damper to close, and then to reconnect the air supply.
This was first done on the exhaust dampers without a problem, even though the exhaust dampers shut momentarily.
When this was done on the dampers on the supply side, the supply dampers went shut and the exhaust fan drew a suction on the Fuel Building apparently great enough to pull the roll-up door off its track and create damage which rendered the fuel building essential ventilation system inoperable.
Problem Resolution Sheet No.
530 was initiated to evaluate the lessons learned from this event and an Incident Investigation Report (IIR) is to be issued documenting the results of this investigation.
Engineering Evaluation Report (EER) 90-ZF-009 has been issued to reevaluate the design basis of the fuel building roll-up door in light of this event:
The inspector will followup on this item when the IIR and EER are complete (528/90-03-03).
No violations of NRC regulations or deviations were identifie Fuel Buildin Radiation Monitor Modi'fication En ineered Safet eatures ctuatlon -
nit After completing a site modification on RU-146, one of the two Fuel Building (FB) ventilation radiation monitors, the technicians re-energized RU-146 and experienced" a high count rate on RU-145 and a
subsequent FB Essential Ventilation Actuation System (FBEVAS) initiation and Control Room Essential Filtration Actuation System (CREFAS)
initiation.
The operators verified that the high count rate was spurious, secured the FBEVAS and CREFAS actuations and emergency lineups, and made the four-hour event report as required by 10 CFR 50.72(b)
(2)(ii).
The high count rate on RU-145 was due to noise generated in the power supply circuits in RU-146.
Similar noise of much less magnitude was observed when this site modification was installed in Unit 2; however, the noise only affected a
second channel in RU-146.
Since the noise did not affect RU-145, it did not cause an Engineered Safety Features (ESF) actuation.
There had been a step in the work order to take RU-145 to bypass as a
precaution; however, it was made "Not Applicable" by the Work Group Supervisor because RU-146 remained in the "Inactive" mode.
It appears that the Work Group Supervisor and Work Control Planner were not aware that noise had.migrated between RU-146 channels at Unit 2.
In addition, the System Engineer was not aware that RU-145 remained on-line.
After the Unit 1 ESF actuation, a Temporary Modification previously installed on Unit 2 was installed in both Units 1 and 3.
The modification was to install a filter to eliminate the noise.
The inspector concluded that communication between the system engineer and Units 1 and 2 Operations, Work Control and I8C could be improved.
No violations of NRC regulations or deviations were identified.
Low Pressure Safet In 'ection (LPSI) "B" Pum Hi h Differential ressure nl While performing American Society of Mechanical.Engineers (ASME)Section XI testing of the "8" Low Pressure Safety Injection (LPSI) pump, the recorded differential pressure (d/p) was 205.75 psid when the Required Action Range was greater than 199.56 psid.
The requirements of IWP-3230, Corrective Action, states that the pump shall be:
(1)
declared inoperative, and (2)
corrected via repair, replacement or analysis or that the test gauges may be recalibrated and the test rerun.
The licensee obtained a different gauge, repeated the test, obtained data which showed the d/p in the Alert Range (197.62 to 199.56 psid)
and declared the pump operable.
The ASME Code requires a baseline d/p to be established and then defines the High Alert Range to be between 2% and 3% above the baseline d/p.
The Code defines the High Required Action Range to be any d/p greater than 3%
above the baseline d/p.
The Code further requires that the gauges used to obtain this data be accurate to at least plus-or-minus 2X of the full scale reading, which shall be three times the reference value or less.
With these requirements, the ASME code permits using gauges which have inaccuracies which are large enough to mask small variations in pump d/p and large enough for the observed value's to read in the Required Action Range with the pump actually operating at the baseline d/p.
Discussion with the NRR Mechanical Engineering Branch advised that it is preferable for the same gauge to be used to establish the baseline and perform periodic testing to eliminate gauge to gauge inaccuracies=.
The inspector concluded that while Code requirements were met, the intent of the Code was to assess the operational readiness of the pump.
The licensee responded by agreeing that their program had a weakness in that different types of gauges were used to establish the baseline and perform the periodic tests.
Licensee management committed to using the same type of gauge for a given pump in the future, but indicated that it would be too difficult to use the same gauge each time.
No violations of NRC regulations or deviations were identified.
15.
'Loss of Instrument Air - Unit j. (93702)
With the unit in Mode 5, instrument air was lost for approximately 46 minutes on February 8, 1990, and approximately 3 minutes on February 9,
1990, due to a failure of an air operated air dryer valve actuator.
The 0-ring in this actuator was worn to the point that it failed to seal.
This caused one tower to remain isolated while the automatic sequencer isolated the other tower resulting in a complete loss of instrument air.
The standby dryer could not be placed in service because it was out of service for maintenance and the nitrogen backup system did not supplement instrument air because it was also out for maintenance.
A relate'd event occurred at Unit 2, while the Unit was at lOOX power, two weeks earlier when the motor which drives the automatic sequencer on the
"A" Train dryer failed.
This resulted in a "High Dew Point" alarm which was masked initially by a failed light bulb.
When the Auxiliary Operator shifted towers, instrument air pressure dropped to 85 psig.
The Auxiliary Operator restored the "A" tower, but before it could supply instrument air loads, the nitrogen backup system supplied the loads, for approximately one minute.
The Auxiliary Operator discovered the "B" tower prefi lter drain trap clogged with "scum" which was discharged when its isolation valve was fully opened.
This blockage allowed a
significant quantity of water to build up in and ahead of the prefilter which blocked "B" Train air flow for a few minutes.
Once this material and water were drained out, the "B" Train functioned properly until the
"A" Train sequencer was repaired.
The compressed gas system has received significant attention from the licensee since problems occurred with the system during a Unit 3 trip on March 3, 1989.
Modifications have been implemented as a result of deficiencies identified following the event as detailed in the updated Technical Report addressing Unit 3 trip concerns, File 89-147-419, dated October 16, 1989.
The corrective actions -have, however, addressed air
quality and the nitrogen backup system and have not focused on the reliability of the instrument air system.
An Engineering Evaluation Report (EER 89-lA-040) was written on October 26, 1989, to address reliability of the instrument air dryers.
This EER was sent to the Nuclear Engineering Department where it received little attention while NED engineers have worked on a design basis review of the, instrument air system in preparation for a presentation of instrument air system modification proposals to the Plant Modification Committee (PMC),
targeted for May 1990.
It is important for the instrument air system, which can have such a
profound impact on operating plant stability, to operate reliably.
The EER written in October 1989, and the two instrument air dryer failures which occurred in January and February 1990, suggest that good reliability has yet to be achieved.
The licensee indicated that the actions on the compressed gas system have been planned in two phases.
Phase one was to improve nitrogen backup system reliability by making certain modifications which have been completed.
Phase two is to improve instrument air system reliability.
This is currently being addressed by the following actions:
(1)
The system engineer has completed an initial review of the operational procedures and written Instruction Change Requests (ICRs) to implement the recommended revisions to the operating procedures.
These ICRs have been approved and the procedures have been updated.
This review wi 11 continue on. an ongoing basis.
(2)
The system engineer has been directed to focus attention on the maintenance issues associated with the instrument air system.
The system engineer has also been directed to keep supervision continually appraised of problems with the instrument air system so engineering management can be involved as necessary to address engineering maintenance issues.
(3)
NED and EED are reviewing the instrument air system and are preparing a presentation for the PMC to recommend modifications to the instrument air system to eliminate performance vulnerabilities and improve system reliability.
This PMC presentation is targeted for May 1990.
No violations of NRC regulations or deviations were identified.
Heated Junction Thermocou le (HJTC) Removal - Unit 1 93702)
As reported in NRC Inspection Report 528/89-54, the licensee experienced difficulty inserting a
new HJTC assembly into the reactor 'vessel following vessel head replacement.
The HJTC in question was one of two assemblies which serve as vessel level indicators during post-accident scenarios when steam/gas voids may exist under the vessel head.
The licensee's investigation determined that a lower section of the HJTC assembly sheathing had broken off of the old assembly during removal and had remained at the bottom of its guide tube, preventing the new assembly from being fully inserted.
The licensee determined that workers removing the old assembly had not noticed the missing sheath.
The licensee's
boroscopic examination within the guide tube initially led them to believe that a welding ridge or burr was responsible for the obstruction and performed some remote grinding and cutting at that location within the guide tube.
Subsequently, the licensee determined the true nature of the problem by. recovering the old assembly from waste containers and boroscopically identifying unique physical characteristics of the stuck portion.
The licensee was then able to remove it using a specially manufactured tool.
The inspector reviewed the engineering documentation and analysis related to this event and concluded the licensee's actions following confirmation of the stuck sheath were methodical, careful, and thorough.
The licensee built -a mockup to assist in their analysis and to experiment with extraction tools, measurement plumb bobs and grinding/cutting tools.
They also obtained timely vendor support and thoroughly. documented all related analyses.
As corrective action, removal procedures for HJTCs are being modified to verify removed assemblies are intact prior to disposal.
The inspector had no further questions.
No violations of NRC regulations or deviations were identified.
Diesel Generator Fuel Oil Stora e Tank Level Meter Inaccurac
- Unit 2 93 02 The Diesel Generator (DG) Fuel Oil Storage Tank (FOST) Level Indicators (LI), LI-33 (for DG "A") and LI-34 (for DG "B"), which are designated as not quality related, have been demonstrated to be unreliable, as documented in licensee maintenance'records.
On February 16, 1990, NRC inspectors noted that LI-33 was indicating that the "A" FOST level was 68K, significantly below the 80K minimum level allowed by Technical Specification (T/S) Limiting Condition for Operation (LCO) 3.8. 1. 1.(b).
After notifying the Shift Supervisor (SS),
an Auxiliary Operator (AO) was dispatched to observe the meter.
When the AO moved'his hand over the face of the meter, the indication moved.
The AO was able to manually adjust the meter reading by this action.
Actual FOST level for both DGs was subsequently determined to be above the minimum requirement by observations of the mechanical indicator in the tank.
This measurement (in feet)
was converted to level by means of data in a book in the Control Room.
Additionally, the FOST low level annunciator, which receives the same level transmitter signal as the meter indication, was not alarming, indicating the level was above 88.5X.
Indicator LI-34, the identical meter on DG "B", was observed to. behave similarly by an AO on December 14, 1989.
On Work Request (MR) 367106, the AO documented that the indicated level had changed from 97K to 794 between the 8:00 PN reading and the 2:00 AM reading, with no explanation.
The low level annunciator was noted to be not alarming.
This MR contained conflicting information regarding whether Train "A" or "B" was faulty, though the AOs logs confirm that the WR was intended to address the "B" meter.
The Assistant SS approved the WR as written and correctly identified that the MR was T/S related.
A "Maintenance Required" tag was hung adjacent to LI-34, describing the observed behavior.
However, the licensee had taken no action as late as February 15,'990.
On the days
prior to submitting the work request, the AO logs show that LI-34 varied substantially between readings without explanation and without being questioned.
The series of successive readings from 8:00 AM, December 13, 1989, to 0:00 AM, December 15, 1989, are (in percent level) 89, 90, 84, 91, 91, 84, 97, 79 and 84, indicating that in spite of these variations and the fact that the readings were not consistent with the state of the level alarm or known plant conditions, operators apparently accepted the reading until it indicated below the low limit documented on the logs.
Although self-verification methods may sometimes be used by operators to determine the appropriateness of readings, it does not appear to be a
consistent practice on routine logs.
The licensee stated that a survey of 25K of all AOs at Unit 2 indicated that all surveyed AOs claim to use self-verification methods routinely.
Five surveillance tests were performed on the "B" DG between December 15, 1989, and February 15, 1990, which used indicator LI-34 to document the FOST level as greater than the minimum T/S value of 80% per LCO 3.8.1.1.b and surveillance requirement-4.8. 1.1.2.a.
Though self-verification methods may have been applied during the performance of the surveillance tests, this was not documented.
Procedure 73AC-9ZZ04, Surveillance Testing in Paragraph 3.2.4. 1 requires that if an indicator is not performing properly or the parameter can only be obtained by the plant computer, that the local indication or plant computer should be used.
On January 25, 1990 an'd four other occasions between December 15, 1989 and February 15, 1990, with indicator LI-34 known to be unreliable, readings were taken from LI-34 for the purpose of satisfying Surveillance Test requirements.
The use of LI-34 in its documented degraded condition appears to be in violation of HRC requirements.
The licensee has committed to revising the surveillance test procedure to use the more reliable local level alarm as the primary indicator of satisfactory FOST level.
Engineering Evaluation Request (EER) 84-DF-005 noted a similar problem in Unit 1 with LI-34 and determined that these indicators, which are d'Arsenval meters, are subject to erroneous indication due to static charges on the plastic face plate.
The EER recommended use of an
'ntistatic spray on the face plate, but this was not incorporated into'ny calibration, maintenance, or surveillance procedure.
Since 1984, the EER program has been strengthened to require initiation of appropriate actions, such as procedure changes, stemming from disposition of EERs.
Following the inspector's questions regarding the "A" meter, Work Order (WO) 409785 was generated to spray an antistatic solution on the face plates of LI-33 and LI-34.
The spray had no observable effect on either indication.
LI-33 was then removed and found to be unresponsive to varying input signals.
After being cleaned internally, and both sides of the face plate sprayed with the antistatic solution, LI-33 was successfully calibrated and reinstalled.
No other cause of the behavior of LI-33 was identified during this maintenance activity.
Because LI-34 appeared to be functioning properly at the time, no further action was taken with it and WO 409785 was closed ou After the exit interview, a Night Order was issued at Unit 2 to direct Auxiliary Operators to alert the Assistant Shift Supervisor if an instrument to be used for a surveillance test has a maintenance required tag on it.
The Assistant Shift Supervisor is to evaluate the tag and decide whether the instrument can be used to perform the surveillance test or not.
If the Assistant Shift Supervisor determines that the meter or gauge can be used, the surveillance test package shall'e annotated with this determination.
One apparent violation of NRC requirement was identified, 50-529/90-03-01.
18.
Pre aration for Refuelin
- Unit 2 (60705)
The Unit 2 Cycle 3 refueling outage preparation was evaluated.
The inspection included a review of selected refueling procedures, attendance at planning meetings and observation of receipt of new fuel.
Procedure 72AC-9NF01, "Control of Special Nuclear Material (SNM) Transfer and Inventory," ensures that Special Nuclear Material is subject to detailed transfer and inventory controls that meet the requirements of 10 CFR Part 70.
The procedure also details the responsibilities of managers, supervisors and engineers in the processing and control of nuclear fuel.
The inspector concluded the procedure established acceptable guidance and controls to maintain accountability and control of special nuclear material.
The inspector's review of 72AC-9NF01 noted that cancelled procedures were referenced.
Specifically, 72MT-9FH01, which had been cancelled and replaced by 31MT-9FN01,
"New Fuel Handling,"
was referenced a number of times.
This indicates a potential problem with procedure administration and attention to detail by Reactor Engineering Management.
The inspector attended outage meetings during which the outage work was statused intra-organizationally.
Work schedules had been developed or were in the process of being developed.
Specific fuel movements and positions were in the process of being determined.
Procedures pertaining to controls over criticality, shutdown margin, containment integrity control, and decay heat removal were written and approved.
The inspector observed the receipt, inspection and storage of new fuel.
These activities were conducted in accordance with Procedure 31MT-9FH01, Revision 0, effective December 7, 1989.
The inspector noted that radiological procedures and controls had been established and were being followed during the receipt of new fuel.
The fuel was examined. for
damage and external contamination.
Material Balance Areas (MBA) had been established to ensure that an inadvertent criticality due to storage configuration was not possible.
Limits had been established as to the number of fuel shipping containers that could be opened and accessed.
The vendor identification number was verified and recorded for each fuel bundle and the new fuel location in which the bundle was stored was recorded.
The inspector noted that the equality Control holdpoints were observed and there was active equality Assurance participation in the receipt, inspection, and storage of the new fuel.
The inspector
'oncluded licensee preparation for refueling was acceptabl.
e-No violations of NRC regulations or deviations were identified.
Ina ro riate
CFR 50.59 Review - Unit 3 (93702)
An approved
CFR 50.59 review was proposed by,the licensee's Engineering Evaluations Department (EED) as authorizing the acceptability of restoring containment vertical tendon V-66 to a tension value less than required by Technical Specifications (T/S).
The licensee's equality Assurance Department, Unit 3 Operations Department, and the NRC Inspector independently determined that it was improper for a 50.59 review to modify a Technical Specification requirement.
Discussion with the EED Supervisor who approved the 50.59 review revealed that his stated intent was to authorize only a temporary condition with'he tendon tensioned to less than that required by T/S.
However, the inspector noted that an approved 50.59 review carries no inherent time limit and, if let stand, would constitute licensee approval of a condition which did not conform with the T/S, which is beyond the authority of a 50.59 review.
The inspector concluded that the EED personnel responsible for performing and approving this 50.59 review did not fully consider the implications of the 50.59 process in authorizing a departure from T/S.
Licensee management acknowledged this concern and stated that the responsible individuals had been counseled.
They also acknowledged the need for improved technical staff training on the 50.59 process.
No violations of NRC regulations or deviations were identified.
CFR Part 21 Pro ram - Units
2 and 3 (36100)
The inspector reviewed the licensee's program for handling the reporting of defects under the requirements of 10 CFR Part 21.
The inspector found that the licensee's program was described in Procedure 94AC-OLC02, Revision 0, "Review of Conditions Adverse to equality for 10 CFR Part 21."
A draft revision to update the procedure was planned for issuance by March 28, 1990.
However, the licensee was implementing the enhanced program at the time of the inspection.
Under the revised program, responsibility for coordinating the Part 21 program has been reassigned to the Compliance Manager.
The inspector found that the licensee program established adequate controls for implementing Part 21 requirements in the following areas:
Posting of required notices, Evaluating deviations, Informing the responsible officer, Notifying the NRC of a reportable defect, Specifying Part 21 applicability on procurement documents, and Maintaining records.
The inspector observed the licensee posted notice in the administration building annex.
The inspector found that the notice appeared to be out of date.
The individual identified on the notice, to contact to report a
deficiency, the Director Corporate equality Assurance/equality Control, was
no longer responsible for coordinating the Part 21 program.
The Compliance Nanager was the appropriate contact.
The inspector identified the out of date posting to the Compliance Hanager who acknowledged the deficiency and initiated appropriate corrective action.
The inspector subsequently observed that an updated notice was promptly posted.
The inspector examined two recent purchase orders, Nos.
60201664 and 33506725, and found them both to contain specifications of 10 CFR Part
requirements.
The inspector examined four recent licensee reports to the NRC of Part
defects; Licensee Event Reports (LER) 89-007-01, 89-0'18-00, 89-004-03 and 89-005-01 and found them to be in accordance with the current program requirements.
The inspector examined two Reportability Evaluation Reports (RERs),
89-07 and 89-13, which the licensee had determined to be not reportable under Part 21 requirements.
The inspector.
found that the licensee's evaluation determination appeared to be adequate and appropriate.
No violations of NRC regulations or deviations were identified.
21.
Alle ation Followu
-
ATS No.
RV-89-A-0030)
a.
Characterization 1)
There is mismanagement of the fitness for duty program.
2)
Independent verification of electrical and instrumentation and control (18C) component status is not always performed during Surveillance Tests (ST) on safety-related equipment and lifting/landing leads is often done without documentation.
3)
4)
Improper 32V fuses are installed in 125V circuits in Inverter'K-12L, which provides electrical isolation between Class lE Control Room annunciator circuits and the Safety Equipment System Status (SESS)
panel in the Control Room.
Al Operability of excess flow check valves in the Emergency Diesel Generator (DG), Essential Cooling Mater (EW), and Auxiliary Feedwater (AF) systems is questionable due to improper testing methodology.
5)
Procedure 36ST-9HP01, Hydrogen Recombiner Operability Test, did not satisfy Technical Specification 3.6.4.2.
6)
The alleger's concerns are not being addressed by management and the alleger is being harassed via management's use of the fitness-for-duty program.
7)
The alleger is being denied access to the unit maintenance manage b.
Im lied Si nificance to Desi n
Construction or 0 eration 2)
The fitness-for-duty program ensures that personnel who have access to nuclear plant systems are not impaired by, drug or alcohol use.
Nismanagement of this program could result in incorrectly identifying an individual as impaired, or could allow an impaired individual access to sensitive plant equipment.
Independent verification of components following testing gives.
added confidence that components are left in a fully operable condition.
3)
4)
5)
6)
7)
Improperly low ratings on fuses could,.result in improper arc quenching and possible re-strike when the fuse blows.
Excess flow check valves are designed to shut:off air or liquid flow from seismic qualified systems to non-seismic portions of those systems in the event of a break in the non-seismic portion.
If excess flow check valves are not operable, the design basis for the system may not be met.
The hydrogen recombiner is required by the plant design basis to eliminate hydrogen produced inside containment following a loss-of-coolant accident.
Operability is required by Technical Specification (T/S) 3. 6. 4. 2.
Hisuse of the fitness-for-duty program can produce a chilling effect on the ability of employee's to raise nuclear safety concerns.
Inaction by management on concerns raised by employees may result in issues significant to nuclear safety not being addressed.
Denial of access by an employee to their management may result in issues significant to nuclear safety not being addressed.
C.
Assessment of Safet Si nificance 2)
Not reviewed in this inspection report.
The inspector reviewed the licensee's equality Investigations Hotline (gIH) Case No. 88-74, Concern No. 8, which addressed-this issue.
The concern was received by gIH on November 29, 1988, from the alleger.
A Corrective Action Report (CAR 88-105)
was initiated on December 21, 1988, to implement an independent verification procedure.
Delayed at least twice, this procedure was implemented on August 28, 1989, as 02AC-OZZOl, "Independent Verification of Valves, Breakers, and-Components."
'Additionally,-the gIH file refers to the biannual review process for all procedures which will include specific chaqges for individual procedures to incorporate independent verification steps where appropriate.
The inspector specifically checked several 18C activities in progress and determined that lifting/landing leads were documented where appropriat )
The inspector reviewed gIH Case No. 88-74, Concern No. 2, which addressed this issue.
The licensee dispositioned Engineering Evaluation Request (EER) 88-ES-004 to replace the fuses with proper 250V rated fuses and issued Work Orders (WO)
No. 341937-Unit 1, 365379-Unit 2 and 365631-Unit 3, in July 1989, to replace the fuses.
The Unit 2 WO was completed on December 31, 1989.
The other two'Os are issued but either waiting for parts or have not yet been completed as of Harch 1, 1990.
4)
5)
The inspector reviewed gIH, Case No. 88-74, Concern No. 22, which addressed this issue.
The licensee determined that EER 86->(N-046 addressed excess flow check valve operability in that as an interim resolution the diesel generator (DG) and auxiliary feedwater (AF) system valves would be isolated except when operators were taking readings on the gages downstream of the check valves.
The inspector previously had identified that the DG system valves remained unisolated following this interim resolution.
This was documented in NRC Inspection Reports 529/89-21 and 529/89-36, and resulted in a violation.
The inspector reviewed the Preventive maintenance (PN) task which tests excess flow check valves and noted that the valve is tested to ensure it closes under specified e'xcess flow conditions.
The "licensee's interim evaluation for the essential cooling water (EW) system was that the excess flow check valves need not be isolated.
Final disposition of EER 86-NH-046 is not complete pending completion of the evaluation for the DG and AF system excess flow check valves.
The inspector reviewed gIH Case No. 89-23, Concern No. 1, which addressed this issue.
The licensee determined that Procedure 36ST-9HP01, Hydrogen Recombiner Operability Test, required revision to include clarification of the acceptance criteria for temperature of operation from "approximately 800 degrees F"
to "greater than 800 degrees F," and the inclusion of appropriate signature blocks for all steps which document acceptance of surveillance test criteria.
The inspector reviewed the licensee's surveillance test procedures and determined that they provided for signature verifications of the surveillance requirements of T/S Surveillance Requirement 4. 6.4.2.
6)
7)
The alleger was informed of,his right to seek redress through the Depar tment of Labor.
In discussions with the NRC inspectors on October 19,'989, the alleger clarified that although he is not being denied access to licensee management, he doesn't feel it does any good to give them his concerns.
Conclusion Not reviewed during this inspectio e 2)
3)
4)
5)
6)
Substantiated.
The inspector concluded the licensee's actions were appropriate, however considerably delayed.
The non-documentation of lifting/landing leads were not substantiated.
Substantiated.
The inspector concluded the licensee's actions were appropriate, however considerably delayed.
Substantiated.
The inspector concluded that the licensee failed to properly address excess flow check valve operability through the EER process.
This is documented in NRC Inspection Report 89-21 and 89-36.
The licensee's action to isolate affected check valves while a final evaluation is being performed appears to be appropriate.
The inspector considered the test methodology to be appropriate.
Partially substantiated.
Clarification to some steps and addition of signature blocks enhanced the procedure to assure that minimum T/S Surveillance requirements were met.
Not Applicable.
7)
The alleger stated that he had access to his management.
e.
Actions Needed for Resolution No further action necessary.
22.
Alle ation Follow-u
-
(ATS No.
RV-89-A-0067)
a.
Characterization Deficiencies identified during post-modification walkdown were not documented.
The deficiencies were corrected without work authorization or documentation of work performance.
Specifically, walkdowns conducted for Modifications SM-IA-007,
=-SM-IA-008 and SM-IA-009 failed to document deficiencies and these deficiencies were corrected without'ork authorization or documentation.
2)
Components (parts)
were procured and intentionally documented as having a equality Class higher than that to which they were procured or documented and as having different critical characteristics than they had.
Specifically, a radioactive drain system 0-ring was evaluated as having a size different than it actually had.
3)
Control room annunciators that were activated (lighted) due to equipment deficiencies, not due to actual alarm conditions, were disabled (jumpered out) by the use of maintenance work orders instead of temporary modifications in order to avoid the additional reviews required by temporary modifications.
4)
Alarm points for the plant computer were changed and or removed without using appropriate administrative controls, including documentation for alerting the plant operator )
Documentation (associated with Item 1 of this Allegation) was intentionally removed from the work bench of an employee to prevent the employee from presenting the. information to the NRC during an announced diagnostic inspection.
b.
Im lied Si nificance to Desi n
Construction or 0 eration 2)
3)
4)
5)
Non-documentation and correction of post modification deficiencies represents a failure to comply with procedures and could result in configuration control problems.
Incorrect classification of procured equipment could result in the failure of a component or system to function as designed.
Disabling annunciators without adequate review could result in annunciators being out of service for extended periods of time without initiating appropriate compensatory measures.
Change or removal of plant computer alarm points could result in inadequate operator response to equipment malfunctions.
Suppression of employees'ights to bring problems forth could create a chilling affect on other employees in addition to potentially not identifying and correcting deficiencies.
c.
Assessment of Safet Si nificance 2)
Site Modifications SM-IA-007, SM-IA-008 and SM-IA-009 installed indication and alarm features in the control room for steam generator nozzle dam air panels=to alert operators to a low nozzle dam pressure.
These modifications were neither safety-related nor quality-related.
However, to evaluate whether or not deficiencies associated with these modifications were being corrected without being, documented, the inspector verified that a sampling of electrical wire terminations were correctly landed.
No deficiencies were noted.
Additionally, the inspector spoke with engineers and craftsmen who installed the modifications to ascertain the administrative controls used to document and correct discrepancies during the implementation of these modifications.
Specifically, the craftsmen and engineers stated all discrepancies had been identified and corrected by a modification to the site modification.
This part of the allegation was not substantiated.
The inspector noted that this concern also had been reported to the,equality Hotline on November 4, 1989, and was in the process of being evaluated.
The inspector interviewed licensee personnel to identify what maintenance had been performed specific to 0-ring replacement on the floor drain system during 1989.
They indicated that prior to 1989, 0-rings for the transmitter cover plate had been inspected as part of Preventative Maintenance (PM) for sump
level transmitters (Magnetron);
and if the 0-rings were found defective, they were replaced.
In 1989, due to a concern about potentially damaging the 0-ring during the inspection process, the licensee decided to always replace the 0-ring during the level transmitter preventative maintenance.
In June, during the performance of the drain transmitter PM, it was identified by a craftsman that one of the transmitter's original (vendor installed) 0-rings was tan, while its replacement was black.
The licensee verified, by contacting the vendor, that the new (replacement)
O-ring, that was in stock, was the correct item by part number.
Additionally, the licensee, through conversation with the transmitter vendor, verified the correct component identification number.
The licensee also indicated that an employee expressed concern (via the equality Hotline) over the difference in color and size (circumference)
between the originally installed and replacement O-ring, and that the evaluation of the concern was nearly complete.
The licensee preliminari ly concluded that the 0-ring originally installed by the vendor was a different color, material and circumference than the replacement O-ring, and the replacement item was the correct size.
= Additionally, the licensee verified that the replacement 0-ring was providing a leak tight seal and properly fit within the recessed slot machined for the 0-ring.
The licensee assigned a low priority to the final evaluation of this concern because the replacement 0-ring provided an effective seal and the vendor verified the replacement 0-ring as the correct item.
The concern over the difference in color and circumference, while real, appeared to be adequately addressed by the licensee.
The licensee is continuing to evaluate this concern to understand if the transmitter vendor either installed an incorrect 0-ring initially or the vendor changed the type and size of the 0-ring after initial manufacturing.
The licensee does not generally measure the circumference of 0-rings on receipt inspection.
The concern over the licensee incorrectly classifying replacement items was not substantiated.
Previously, the NRC has expressed concern with the licensee over the apparent high number of inappropriately activated annunciators in the control room.
Additionally, the NRC had expressed a concern over the apparent high number and duration of facility temporary modifications.
It appeared that the temporary modifications to systems and components were not being addressed in a timely manner and were, therefore, essentially permanent design changes that in some cases altered the original design.
To address these concerns, the licensee implemented a tracking list for "inappropriately lit annunciators 'nd "appropriately lit annunciators."
The tr acking list is evaluated on a daily basis by the shift supervisors to ensure daily work activities
4)
for annunciators are being actively pursued.
Additionally, the Plant Nanager reviews the tracking list with station maintenance and engineering managers at a weekly planning meeting to assure work is planned to correct annunciator deficiencies.
The licensee also establishes compensat'ory measures that are required to appropriately monitor plant parameters when annunciators are deactivated.
The licensee, in discussions with the inspectors, stated that, in general, work orders are generated to correct deficiencies with annunciators
.vice using temporary modifications because the deficiency can be evaluated and corrected rapidly. If the deficiency is a result of a design inadequacy, then a temporary modification is generally generated.
The current licensee procedures permit either method of correction and is silent on preference of technique.'he allegation was not substantiated.
The licensee continues to have a relatively high number of inactive annunciators; however, the number of inactive annunciators is trending down.
The inspector noted that the concern of removing plant computer alarm points was being tracked by the Employee Concerns Program (ECP)
as item 89-23-07.
Licensee investigation of the concern was in progress.
The licensee had, to date, identified deficiencies in the administration of changes to plant computer alarms; however, alarm setpoints are entered by licensed operators.
Therefore, plant computer alarms could have been removed without the operator's knowledge; however, setpoints would not have been changed without operator recognition.
Based on the plant computer being nonsafety-related equipment and the licensee continuing followup on this employee concern, this item is closed.
5)
The NRC inspector found that documents had been removed from an employee's desk after the employee had terminated his employment at Palo Verde.
The materials (including documents)
that the employee left behind were collected and retained by the licensee.
The NRC inspector accompanied by a licensee representative, examined the materials that the employee left behind and could not find documentation associated with item 1 of this allegation.
The licensee was tracking this issue in the Employee Concerns Program as item ECP-89-105-11.
The licensee's investigation did not substantiate this allegation.
This item was not substantiated.
d.
Conclusion 1)
2)
3)
4)
5)
Not substantiated.
Not substantiated.
Not substantiated.
Partially substantiated.
Not substantiate e.
Actions Needed for Resolution N'one.
This allegation is considered closed.
No violations of NRC regulations or deviations were identified.
23.
Alle ation Follow-u
-
ATS No.
RV-89-A-0068)
a.
Characterization, The inspector reviewed an allegation involving management qualifications and engineering changes.
This review addressed three areas of potential concern:
1)
qualifications of APS managers and supervisors.
2)
Plant changes identified by Engineering as being required resulting from Information Notice (IN) 86-29, Information Bulletin (IB) 85-03 and Significant Operating Experience Report (SOER) 86-02 have not been accomplished.
3)
Corrective actions resulting from the failure of two motor operated valves during the event of July 6, 1988, have not been accomplished in Unit 2.
b.
Im lied Si nificance to Desi n
Construction or 0 eration 1)
Hang ement ualifications Paragraph 13. 1.3 of the Updated Final Safety Analysis Report (FSAR), describes that personnel in certain positions must meet certain minimum qualifications as described in ANSI/ANS 3. 1-1978.
The combination of education, experience and skil.l commensurate with the level of responsibility is intended to provide reasonable assurance that decisions and actions during normal and abnormal plant conditions will be such that the plant is operated in a safe and efficient manner.
Personnel who do not meet the minimum selection criteria may not be able to properly direct the activities of operation and.
support organizations, potentially affecting the operation of the plant.
2)
Re uired Plant Chan es The plant changes referred to in the allegation concern the rewiring of the motor operators of safety-related valves to actuate the torque bypass swit'ch from a separate rotor in the geared limit switch assembly.
This modification had been identified in the referenced documents as a recommended modification to preclude recurrence of operational problems which had resulted from the conventional use of a common rotor
'
l,i
3)
in the limit switch assembly for both valve position indication and motor control functions.
Failure to implement the modification could result in the failure of butterfly valves which are adjusted with minimal torque switch bypass to operate properly.
Also, inaccurate valve closed position indication could result for throttle valves, providing misleading information to the operators.
Corrective Actions for Unit 2 The change referred to in the allegation concerns corrective actions taken in Units 1 and 3 which allegedly were not accomplished in Unit 2.
The changes were identified to preclude recurrence of the operational problems experienced on two butterfly valves during the Unit 1 auxiliary transformer fire on July 6, 1988.
The change involves implementing the rewiring modification described in (2) above and increasing the setting of the torque bypass switch to allow additional valve travel during opening before enabling the torque switch for overload protection.
Failure to implement the identified corrective actions could result in additional failures of similar valves in Unit 2 when required to open under off-normal operating conditions.
However, the safety function of the valve to close would not be affected by the deficiency.
c.
Assessment of Safet Si nificance Mana ement ualifications The inspector reviewed the records of the qualifications for various managers.
The inspector found the licensee's records to consist solely of the personal resume submitted by the individuals for employment.
The 'inspector found no record of a licensee evaluation of the qualifications stated in the resumes to determine whether the combination of education and experience satisfied the minimum criteria for positions under ANSI/ANS 3. 1-78.
With the exception of one person, the inspector found that records of the nine managers and supervisors reviewed appeared to satisfy the minimum selection criteria for their functional level as described in ANSI/ANS 3. 1.
The inspector found that the resume for one individual did not appear to meet the minimum qualifications under Paragraph 4.4.5 of ANSI 3. 1.
According to licensee personnel, that individual held the position for approximately 'one year until reassigned to another position.
The individual is no longer employed by APS.
The lack of qualifying experience for the individual was previously addressed by NRR in an Safety Evaluation Report (SER) dated October 2, 1987.
NRR found the qualifications to be acceptable
based on additional training as discussed in Inspection Reports 87-44, 88-29 and 89-16.
Weaknesses in the documentation of personnel qualifications and training have been previously identified by the licensee in gA audits '88-11, 89-11, 89-23 and Corrective Action Report (CAR)
88-54.
The inspector reviewed licensee procedures 73DP-OTR02 Revision 0, "qualification and Training Requirements for System Engineers",
and 15AC-OTR01 Revision 1, "Personnel qualification and Certification."
The -inspector found that the new procedures appeared to incorporate adequate clarification and documentation of personnel qualification evaluations.
Re uired Plant Chan es The inspector reviewed the actions taken by the licensee in response to IN 86-29, IB 85-03, and SOER 86-02.
The inspector found that the original Independent Safety Engineering Group (ISEG) review of SOER 86-02 was closed based on evaluations by Operations Engineering which concluded that no actions were required.
Th>s conclusion was based on the understanding by Engineering that four rotor limit switches were already installed in all safety-related MOVs during startup.
However, this review did not identify that wiring changes to relocate the torque bypass switch to a separate rotor in the limit switch assembly had'not been made at the time that the four rotor assemblies were installed.
The licensee review was subsequently reopened as a result of an Institute for Nuclear Power Operations (INPO) evaluation.
A Rotor Usage Review was subsequently conducted and an Engineering Analysis Request (EAR) 89-1744 was initiated to develop the rewire modification packages for the safety-and nonsafety-related MOVs affected.
Implementation of the rewire modification was scheduled in Unit 2 during the upcoming refueling outage and subsequently in Units 1 and 3 during their next refueling outage to be completed by June 30, 1991.
The inspector found that twenty. Design Change Packages/Plant Change Packages (DCP/PCP)
had been generated by Engineering to accomplish the modification.
The inspector reviewed the outage plan for Unit 2 and found that 85% (217) of the affected MOVs were planned to be modified during the outage.
According to licensee representatives, the remaining 15K (37 MOVs) were under review for potential inclusion within the scope of the refueling outage if not accessible for the modificati'on while at power.
The inspector found that the licensee considered the modification to be an improvement in the reliability of the equipment.
Interim operability of the valves without the modification was considered acceptable based on the existing surveillance testing.
Based on the scope of work identified and scheduled for Unit 2, the inspector concluded that the required modifications appeared to be in progress and adequately controlled to
accomplish the modifications within the licensee's committed schedule.
3)
No specific action was required by IN 86-29.
The licensee actions with regard to IB 85-03 were inspected in NRC inspection report 50-528/88-19 and found to be appropriate.
Corrective Actions for Unit 2 The inspector reviewed the scope of work planned for the upcoming Unit 2 outage.
The inspector found that valves EW-65 and EW-145 were scheduled to be rewired per DCP-2FJEW-032 to incorporate the modifications which had been previously performed in Units 1 and 3.
The inspector concluded th'at equivalent corrective actions were scheduled for Unit 2 ~
The licensee actions resulting from the failure of the two valves during the fire at Unit 1 in July, 1988, were previously reviewed in Inspection Report 528/89-34 and found to be adequate.
d.
Conclusion Mana ement ualifications The inspector found that this item was not substantiated.
The personnel currently assigned in the management positions were found to meet the minimum 'selection criteria of ANSI/ANS 3. 1-78.
2)
3)-
Re uired Plant Chan es The inspector found this item partially substantiated.
Implementation of the required MOV rewiring had not yet been completed; however, the modification appeared to be adequately scheduled for completion in Unit 2 during the upcoming outage and in all Units by June, 1991.
Unit 2 Chan es The inspector found that this item was partially substantiated.
The corrective actions required had not been accomplished in Unit 2.
However, as discussed in Item 2 above, the required modifications were found to be appropriately scheduled for completion during the upcoming outage.
e.
Actions Needed for Resolution None.
This allegation is considered to be close.
Alle ation Followu
- (ATS No.
RV-89-A-0069)
~
~
~
a.
Characterization 1)
Inadequate disposition of Corrective Action Reports (CAR).
2)
A design change was accomplished in all three units using only work orders, (WO 234012 Unit 1, WO 287345 - Unit 2, and WO 365597 - Unst 3) raising concerns regarding configuration control.
b.
Im lied Si nificance to Desi n
Construction or 0 eration j.)
Problems identified by plant personnel must receive appropriate action by management.
Failure to do so could have substantial negative effects over the life of the plant.
These CARs are significant in that operational errors may result from poor status control.
2)
Engineering considerations may be being circumvented by not using formal design change procedures for plant modification.
In bypassing these procedures, procedures and drawings may not be updated to reflect the modifications and material compatibility may not be considered.
c.
Assessment of Safet Si nificance 1)
Adequate disposition of the referenced CARs was achieved to the satisfaction of both the inspector and the alleger.
The alleger had the opportunity to raise any concerns regarding the disposition.
Therefore, this item has limited safety significance.
2)
The specific design change was ultimately made into a site modification (SH-SR-011)
using appropriate procedures.
However, the interim conditions existed for about two ye'ars.
Corrective actions have been implemented to prevent recurrence.
The licensee's decision to perform the work via work orders instead of via design change procedures was based on expediency and a lack of responsiveness of the engineering organization to operational needs.
As a result of pressure from the licensee's equality Assurance organization, action was taken to address the modification by the approved process.
d.
Staff Position 1)
The item was not substantiated.
2)
The item was substantiated, but corrective action was taken to prevent recurrence.
A~t.i R
R d
Non $
This allegation is considered closed.
Review of Licensee Event Re orts - Units
2 and 3 (90712 and 92700)
The following LERs were reviewed by the Resident Inspectors.
528/88-10-Ll (Closed)
"Ground Fault in 13.8KV Bus Causes Fire sn Unit Auxiliar Transformer and Reactor Trl
- Unit l.
The licensee issued this LER Supplement to document the results of the engineering evaluation and to clarify the power supply arrangement for the deluge system.
The corrective actions have not changed and are complete.
This LER is closed.
No violations of NRC regulations or deviations were identified.
528/89-03-LO (Closed)
"Loss of Power to Alternate Fuel Bui ldin uen a
sa
>on onstor -
ni This licensee event describes an apparent unmonitored effluent release that occurred for up to two hours and 55 minutes on February 17, 1989, from the Fuel Building (FB).
On February 17, 1989, the licensee installed a temporary portable radiation monitor Preplanned Alternate Sampling Points (PASP)
on the FB exhaust ventilation system because the installed radiation monitors, RU-145 and RU-146, had been declared inoperable due to their fai lure during a source check.
On February 17, 1989, at 2: 15 PM, a chemistry technician, who was verifying adequate flow rate, identified that the PASP monitor had de-energized.
At approximately 2:45 PM, the PASP monitor was returned to service.
As corrective action, the licensee identified electrical distribution circuits that were near maximum loading and generated a Plant Change Request to supply dedicated power to PASP monitors.
Additionally, the licensee stated that verbal direction had been given to technicians to verify, in the future, that circuits were not overloaded prior to energizing the PASP monitors and that tagging of the receptacles had occur red that note the receptacle to be dedicated to the PASP system.
The inspector's evaluation of this event included assessing the timeliness and quality of the report, a review of the proposed corrective actions and a review of plant change package (PCP) 88-01-ZZ-08, that will install the dedicated power supply for the PASP monitors.
The inspector noted that LER 89-03 slightly exceeded the 30 day reporting requirement, three days overdue.
The inspector concluded the proposed plant change, when installed, should prevent loss of PASP monitors due to inadvertent de-energization of the monitor or overloading the monitor's electrical supply circuitry.
Because the installation of the change will not occur in February 1990, the licensee committed to submit a supplemental LER.
The inspector verified that the licensee had complied with Technical Specification 3.3.3.8 actions 36, 37, 41 and 42.
To prevent violating these Technical Specifications due to an inoperable permanently installed
t l0'
'2
'
radiation monitor, the licensee requires taking a sample every two hours and verifies flow.
The sample and flow verification are tracked on a
followup report (FUR).
FUR 1-89-0009 was generated for the inoperability of radiation monitors RU-145 and RU-146.
e 26.
The L'ER is closed based on licensee generation of PCP-01-22-08.
528/89-07-LO/L1/L2 (Closed
"Pressurizer Safet Relief Valve Set-Points u
o o erance -
n>
On April 12, 1989, licensee personnel in Unit 1 discovered that two of the four pressurizer code safety relief valves did not meet the Technical Specification (T/S) setting requirements of 2500 psia plus-or-minus one percent.
The plant was in Mode 4 at the time and two code safety valves were operable when only one was required.
The licensee adjusted and retested the two inoperable valves and declared them operable.
The licensee has performed a root cause of failure engineering evaluation, which has resulted in a determination that the code safety valves have a
performance limitation for setpoint repeatabi lity of only plus-or-minus three percent.
LER Supplement 2 indicated that the licensee is pursuing a T/S change to relax the current requirements.
The licensee acknowledged that the existing T/Ss will be met until a license change is approved.
This LER is closed.
528/89-13-LO (Closed)
"Potentiall Un uglified Containment Pur e
so at>on a ves -
nit On July 26, 1989, the licensee reported that the hand wheels on the eight inch containment power access purge valves were unqualified and that these valves may not operate properly during or after a seismic event.
The hand wheels were removed at all three units and an evaluation was performed to provide an ability to manually operate these valves.
The design has been approved by the Plant Modification Committee and will be implemented in September 1991.
A supplement to this LER will be issued soon to document the results of the incident investigation report.
This LER is closed.
Review of Periodic and S ecial Re orts - Units
2 and 3 (90713)
Periodic and special reports submitted by the licensee pursuant to Technical Specifications (T/S) 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 validity of the reported information.
Within the scope of the above, the following reports were reviewed by the inspector.
Unit 1 o
Monthly Operating Report for January 199 )
e 3 Unit 2 o
Monthly Operating Report for January 1990.
Unit 3 o
Monthly Operating Report for January 1990.
No violations of NRC requirements or deviations were identified.
27.
Exit Meetin (30703)
The inspector met with licensee management representatives periodically during the inspection and held an exit meeting on March 1, 199 'J I