IR 05000528/1989016
| ML17304B229 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/23/1989 |
| From: | Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17304B225 | List: |
| References | |
| 50-528-89-16, 50-529-89-16, 50-530-89-16, NUDOCS 8906090113 | |
| Download: ML17304B229 (44) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION V
Re ort Nos.
Docket Nos.
License Nos.
Licensee:
50-528/89-16, 50-529/89-16 and 50-530/89-16 50-528, 50-529, 50-530 NPF-41, NPF-51, NPF-74 Arizona Nuclear Power Project P.
0.
Box 52034 Phoenix, AZ. 85072-2034 Ins ection Conducted:
March 20, through April 26, 1989.
Inspectors:
T. Polich, Senior Resident Inspector D.
Coe, Resident Inspector G.
i r lli, si nt Inspector Approved By:
Miller, ef Reactor Projects Branch,Section II 5 -zg-g Date Signed Ins ection Summar Ins ection on March 20 throu h
A ri 1
1989.
(Re ort Nos.
50-528/89-
50-529/89-16 and 50-530/89-16 Areas Ins ected:
Routine, onsite, regular and backshift inspection by the three resident inspectors.
Areas inspected included: previously identified items; review of plant activities; engineered safety feature system walkdowns; monthly surveillance testing; monthly plant maintenance; engineered safety feature'ystem walkdowns Units 1, 2 and 3; monthly surveillance testing - Units 1, 2 and 3; monthly plant maintenance
- Units 1, 2 and 3; containment local leak rate testing Units 1, 2, and 3; inoperable atmospheric dump valve No.
178 - Unit 1; improper installation of emergency diesel generator fuel line Unit 1; multiple electrical load sheds
- Units 1, and 2; corroded diesel generator intercooler elbow - Unit 2; mid-loop operations
- Unit 3; 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:
30703, 40500, 61720, 61726, 62703, 71707, 71710, 90712, and 93702.
Safet Issues Mana ement S stem (SIMS Items:
None 8906090ii3 890526 PDR ADOCK 05000528
PNU
Results:
Of the seven a'reas inspected, two violations were identified.
These violations pertained to failure to follow procedures and inadequate procedures.
Load sheds of the NAN-SOl bus occurred on Units 1 and 2 occurred several weeks apart.
It did not appear that lessons learned from the first load shed were adequately conveyed from Unit 1 to the other units.
The corrective action for a previous corrosion problem of drain plugs on diesel generators was insufficient in that an elbow fitting apparently failed due to the same mechanism.
General Conclusions and S ecific Findin s
Si nificant Safet Matters:
None Summar of Violations:
Summar of Oeviations:
None 0 en Items Summar 2 items closed, 1 item left open, and 9 new items opene DETAILS 1.
Persons Contacted:
The below listed t'echnical and supervisory personnel were among those contacted:
Arizona Nuclear Power Pro ect ANPP
"R. Adney,
"J. Allen, J. Bailey, B. Ballard, P. Brandjes, C.
Churchman, W.
Conway,
"J.
Haynes, D. Heinicke, P. Hughes,,
"W. Ide, D. Karner,
"J. Kirby, J.
LoCicero, W. Marsh, A. McCabe, D. Phil 1 ips, J. Reilly,
- C. Russo,
"T. Shriver, G.
Sowers,
"R. Younger,
"W. Quinn, Plant Manager, Unit 3 Relief Plant Manager Assistant. Plant Manager Unit 3 Quality Assurance Director Central Maintenance Manager Work Control Manager, Unit 3 Executive Vice President
- Nuclear Vice President, Nuclear Production/Site Director Plant Manager, Unit 2 Radiation Protection
& Chemistry Manager Plant Manager, Unit 1 Vice President - Nuclear Director, Nuclear Production Support Independent Safety Engineering Manager Plant Director Maintenance Manager, Unit 1 Outage Management Manager Standards and Technical Support Director Assistant Quality Assurance Director Compliance Manager Engineering Evaluations Manager Plant Standards and Control Manager Nuclear Safety and Licensing Director The inspectors also talked with other licensee and contractor personnel during the course of the inspection.
~Attended the Exit meeting held with NRC Resident Inspectors on April 26, 1989.
2.
Previousl Identified Items - Units 1
and
92702 92701 a.
(Closed Followu Item 528/529/and 530/87-44-01:
"Trainin for ualit Assurance A
Director" Units 1 2 and 3.
This matter dealt with a licensee commitment to the Nuclear Regulatory'ommission involving the previous QA manager who was to undertake special training while holding that position.
This commitment no longer applies since the individual was recently reassigned to another management position.
This item is close b.
0 en Followu Item 529/88-42-01:
"Bushin s for En ineered Safet Features ESF Service Transformers"
- Unit 2.
The item was that there were two sizes of bushings (B0-5 and BO-6) for the ESF auxiliary transformer at the Palo Verde Nuclear Generating Station.
These two sizes of bushings are not interchangeable as spare parts.
Only one size (BO-6) was
.
on hand in the warehouse as spare bushings when in early January 1989, the Unit 2 ESF auxiliary transformer experienced damage to a bushing (size B0-5).
The replacement bushing was finally found in an APS warehouse in Phoenix.
To preclude this in the future, the licensee initiated EER 89-NB-001 dated February 1,
1989 to inspect the bushings of all ESF auxiliary transformers, during unit refueling outages to determine the type and number of spare bushings, and to stock them in the warehouse.
This item remains open pending the final dispositioning of EER 89-NB-001.
C.
Closed Fol 1owu Item 530/88-41-01:
"Entr Into Mode
Cold Shutdown
" - Unit 3.
This item was opened following the inspector's observation that procedure 74AC-9CY04, "System Chemistry Specification" allowed a waiver of RCS hydrogen and total gas limits prior 'to entering Mode 5 due to a Limiting Condition of Operation, but did not allow a similar waiver for the hydrogen limit prior to securing Reactor Coolant Pumps (RCPs).
This gave rise to the appearance of a procedural violation when RCP's were secured in accordance with the Mode 5 entry procedure, and Plant Manager authorization to remain above the specified hydrogen limit was not obtained.
The inspector reviewed the licensee's Procedure Change Notice No.
03 to 74AC-9CY04 and determined that the requirement to obtain management approval when waiving hydrogen and total gas limits was clarified.
Also, the basis for these limits was stated, and an evaluation of the effects of the waiver on these bases was additionally requi'red.,
The inspector concluded that the procedure adequately stated the nature and extent of authorization required to waive these limits.
This item is closed.
3.
Review of Plant Activities 71707 71710 93702 a ~
Unit 1 Unit 1 remained shutdown during the report period.
Due to the decision to fully investigate the failures of the Atmospheric Dump Valves (ADVs) at Unit 1 and 3, utility management decided to commence its second refueling outage.
Mode 5 was entered on April 14, 1989.
The outage was expected to last 87 days with Mode 2 entry scheduled for July 1, 1987.
On April 23 an Unusual Event was declared due to loss of communication capability due to a power outag b.
Unit 2 Unit 2 remained in Hot Shutdown (Mode 3) during the report period.
Atmospheric Dump Valve testing was conducted and two ADVs were inoperable at the end of the report period.
The Unit 2 staff and other support organizations participated in two emergency preparedness practice exercises in preparation for the May 3, 1989 Emergency Drill.
C.
Unit 3 Unit 3 remained shutdown for a refueling outage during this report period.
At the beginning of the report period, Unit 3 was in Mode 5.
Mid-loop operations were conducted for approximately 12 days during this period.
Outage work was performed, including Reactor Coolant Pump (RCP) removal, reactor vessel head removal, core defueling, and steam generator eddy current testing.
At the end of the report period, Unit 3 was in refueling mode with the entire core off-loaded to the spent fuel pool.
d.
Plant Tours The following plant areas at Units 1, 2 and 3 were toured by the inspector during the inspection:
Auxiliary Building Containment Building Control Complex Building Diesel Generator Building Radwaste Building Technical Support Center Turbine Building Yard Area and Perimeter The following areas were observed during the tours:
l.
0 eratin Lo s and Records Records were reviewed against Technical Specification and administrative control procedure requirements.
2.
Monitorin Instrumentation Proces's instruments were observed for correlation between channels and for conformance with Technical Specification requirements.
observed for conformance with 10 CFR 50. 54.(k), Technical Specifications, and administrative procedures.
4.
E ui ment Lineu s
Various 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 the conduct of partial system lineups.
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 systems from fulfillingtheir functional requirements.
7.
Fire Protection Fire fighting equipment and controls were observed for conformance with Technical Specifications and administrative procedures.
8.
Plant Chemistr Chemical analysis results were reviewed for conformance with Technical Specifications and admin-istrative 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.
10.
Pl ant Housekee in Pl ant condi tions and material/equipment storage were observed to determine the general state of cleanliness and housekeeping.
Housekeeping in the radiologically controlled areas was evaluated with respect to controlling the spread of surface and airborne contamination.
ll.
Radiation Protection Controls Areas observed included control point operation, records of licensee'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 Feature S stem Walkdowns - Units 1 2 and
(71710 Selected engineered safety feature 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 the walkdown of the systems, items such as hangers, supports, electrical cabinets and cables, were inspected to determine that they were operable, and in a condition to perform their required functions.
Accessible portions of the following systems were walked down during this inspection perio Unit 1 o
"A" and "8" Trains.
o Control Room Essential Filtration System
"A" Train.
Unit 2 o
Hydrogen Analyzer "8" Train.
o Auxiliary Feedwater System
"A" Train.
Unit 3 o
"8" Train.
No violations of NRC requirements or deviations were identified.
5.
Monthl Surveillance Testin Units
2 and 3 (61726 Selected 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.
b.
Specifically, portions of the following surveillances were observed by the inspector during this inspection period:
Unit 1 Procedure Descri tion o 41ST-1DG01 Diesel Generator
"A" Test.
o 73TI-9SG05 Atmospheric Dump Valve Functional Test - Valve Nos.
178, 179, and 185.
o 41ST-1SG03 Testing ADVs in Mode 3.
Unit 2 Procedure Descri tion o 42ST-2AF02 Auxiliary Feedwater Pump AFA-POl Operability Test.
o 72ST-2RX09 Shutdown Margi Unit 3 o 74ST-9ZZOl Refueling Boron Surveillance Test o 73ST-9ZZ22 Mechanical Snubber Functional Test No violations of NRC requirements or deviations were identified.
Monthl Plant Maintenance
- Units 1 2 and
62703 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 gA/gC 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.
Specifically, the inspector witnessed portions of the following maintenance activities:
Unit 1 Descri tion o
"B" Train Emergency Diesel Generator Preventative Maintenance
.
o Nitrogen Regulator Removal from No.
184 Atmospheric Dump Valve Supply Line.
Unit 2 Descri tion o
No.
178 Atmospheric Dump Valve Accumulator Blowdown for Particle Measurement.
o No.
184 Atmospheric Dump Valve Accumulator Blowdown for Particle Measurement.
o Disassembly, Cleaning, No.
178 Atmospheric Dump Valve Nitrogen Relief Valve.
Unit 3 o
Overhaul Emergency Diesel Generator
"A".
II
o Overhaul of Main Steam Isolation Valves Control Valves.
Removal of No.
178 Atmospheric Dump Valve; Reactor Coolant Pump 2A Disassembly.
o Disassembly and Inspection of No.
184 Atmospheric Dump Valve.
7.
Containment Local Leak Rate Testin
- Units 1
and
61720 The inspector reviewed the licensee's Local Leak Rate Testing (LLRT)
program in the following areas:
Personnel qualification to ANSI/ANS 3. 1-1978,
"Standards for Selection and Training of Personnel for Nuclear Power Plants."
Personnel training Test instrument calibration Test methodology The LLRT Engineer and Lead Engineer were both employed by ANPP and met ANSI/ANS 3. 1-1978 qualification requirements for Staff Specialist (Section 4.7.2)
and Engineer in Charge (Section 4.6. 1),
respectively.
Both individuals met Section 4.3.2 requirements for supervisors not requiring NRC licenses.
This conclusion was drawn from a review of the training file for each individual.
Experience in lieu of a Bachelor's Degree was used as provided by the standard for one applicable individual.
In addition, the inspector verified that LLRT technicians were qualified in accordance with Section-4.5.2 or were participating in a training program per Section 5.3.4.
In addition, five contractors were employed as LLRT Field Supervisors during the current refueling outages and documentation of their qualifications were also reviewed and found to meet Section 4.3.2 requirements.
The inspector noted that an LLRT training program had been established since January 1988, in support of meeting ANSI/ANS 3. 1-1978 requirements for LLRT personnel.
However, the inspector noted that one technician, who was certified under this program to perform several LLRT procedures, was decertified on these procedures only 13 days after certification.
Supervisors for LLRT determined that decertification was warranted based on their assessment that the technician's knowledge level was inadequate.
The same supervisor who certified the technician subsequently withdrew the certifications.
Because of the relatively short period of certification, the inspector concluded that the initial evaluation of.the technicians'nowledge and ability was inadequate for the purpose of determining initial certification.
The inspector commented to licensee management that there appears to be a need to improve the evaluation capability of the LLRT certification program.
The licensee acknowledged the inspector's comments and agreed to review the matter furthe The inspector noted that, in August 1988, the licensee identified the need to correct air flow measurement device (rotameter)
raw readings to account for the use of test pressures different from the rotameter calibration pressure and the use of test fluids such as nitrogen, different from the fluid used for calibration (air).
The inspector reviewed the licensee's Engineering Evaluation Request (EER) 88-CL-04 dated August 23, 1988, which assessed previous LLRT results against acceptance criteria corrected for the difference between test and calibration pressures.
This EER identified ten containment airlock seal LLRT results which did not meet the corrected acceptance criteria.
The inspector reviewed EER 88-CL-05, dated September 16, 1988, which identified additional conservatisms which had not been previously applied to the LLRT method used for airlock seal tests such that the ten unsatisfactory results were then found to be satisfactory when compared to the corrected acceptance criteria.
The inspector concluded that the licensee had established new LLRT acceptance criteria based on well understood correction factors, had reviewed.past LLRT results against the new criteria, and had resolved discrepancies using appropriate technical justification.
The inspector noted that once these new criteria were established, it took nearly four months to formally change the surveillance test procedure and that, during the interim, LLRT results were being evaluated against an internal memorandum which promulgated the new criteria.
The licensee's procedure 01AC-OAPOl,
"Format and Content of Nuclear Administrative and Technical Procedures,"
Section 3.4.2, requires that changes to procedures be reviewed and approved in accordance with procedure 01AC-OAP02,
"Review and Approval of Nuclear Administrative and Technical Procedures".
The inspector considered that the implementation of new criteria directed by the memorandum constituted a change to the surveillance test procedure, and that between September 1, 1988, (the date of the memorandum)
and December 23, 1988, (the date of the Procedure Change Notice) this change was implemented without benefit of the required review or approval.
This is considered a violation of regulatory requirements (528/89-16-01).
The inspector reviewed the licensee's inflow and outflow test methodologies.
He noted that the LLRT surveillance procedures did not explicitly show that when the outflow method was used, it was equivalent to or more conservative than the inflow method.
This determination prior to using the outflow method is recommended by ANSI/ANS 56.8-1987,
"Containment System Leakage Testing Require ments,".
Paragraph 16.5.2.
For testing done on non-pressurized systems, the outflow method will not detect boundary leakage which escapes through paths other than via the flow detector.
The inflow method, however, conservatively detects non-boundary leakage as well as boundary leakage.
Since most LLRT testing done at Palo Verde utilizes the outflow method, and the licensee's FSAR appears to implicitly allow only the inflow me'thod, the inspector determined that additional review was necessary to establish the adequacy of the licensee's procedures.
This item is open pending further review (528/89-16-02).
8.
Ino erable Atmos heric Dum Valve ADV No.
178 Unit 1 (71707 On April 10, 1989, a reactor operator, in support of maintenance activities was directing an auxiliary operator in performing valve manipulations on the nitrogen system for ADVs.
The reactor operator noticed that the valve status prints indicated that the nitrogen supply valve SGE-V354 to ADV No.
178 was closed.
The reactor operator, knowing that the only ADV operable was ADV No.
178, directed the auxiliary operator to verify the position of SGE-V354
~
The auxiliary operator reported back that SGE-V354 was closed.
The reactor operator informed the Assistant Shift Supervisor of the problem and the valve was opened.
An Operating Department Guideline (ODG) No. 17, "Valve/Electrical Alignment", was performed for 1SGB-HV-178,.and an independent verification was performed.
This action returned the ADV to an operable condition, restoring compliance with the requirements of the Technical Specifications, which required an operable ADV in this mode.
Atmospheric Dump Valve ADV No.
178 had been declared operable on April 6, 1989, following performance of surveillance test 41ST-1SG03,
"Testing ADVs In Mode 3".
According to the licensee,
'between the declaration of operability of ADV No.
178 on April 6, and April 10, the only known activities involving ADV No.
178 were the daily nitrogen accumulator pressure checks.
No specific cause for the closure of SGE-V354 could be determined, however, the investigation of this event by the licensee produced the following findings:
1.
The nitrogen isolation valve, SGE-V354, was found closed, rendering ADV No.
178 inoperable.
2.
Appendix "M" of procedure 410P-1SG01, Revision 8, "Main Steam" which contained the operating instruction of the ADVs and was posted at the valve locations, was deficient in that it did not contain independent verification of critical valves in the system.
3.
The operating crews did not recognize that a low ADV accumulator pressure alarm would be received when instrument air was isolated for work on a compressor ADV.
This was an expected phenomenon due to the inherent leakage in the valve positioner and I/P converter.
Had they been sensitive to this they would have realized a problem existed with nitrogen supply to the ADV.
4.
The nitrogen supply valves for the ADV accumulators required an equalization of pressure across the valve to permit the valve to open.
The valve handwheel could indicate the valve was open when it was actually closed.
o While this was not determined.to have contributed to the valve closure, this valve design does provide the
t
potential for the supply valve to be closed when it indicates open.
The Licensee's corrective actions included the following:
o Procedure changes to clarify instructions and provide independent verification of valve position.
o Addition of valves in the nitrogen and air supply system to the locked valve/breaker control procedure.
o Provide training for operators to highlight the fact that low accumulator alarms are expected when instrument air is isolated to ADVs.
o Conduct a
human factors investigation of the performance of operating staff associated with the event.
o Conduct an engineering evaluation to determine if a more suitable nitrogen supply valve can be used to eliminate the potential for the valve to be closed when thought to be open.
The inspector concluded that the licensee appeared to be considering the appropriate concerns associated with this event.
At the time of the event, the plant was in Mode 3.
Had a condition existed where plant cooldown was necessary, the ADVs could have been operated manually.
This is an undesirable option but was still available with the main steam safety valves also available for the release of steam to the atmosphere had steam pressure exceeded the setpoints of the safety valves.
Operation of the atmospheric dump valves is governed by operating procedure 410P-lSG01, Revision 8, "Main Steam".
A prerequisite in Section 4.2 of the procedure for placing the main steam line in service with the main steam isolation valves open is to complete the valve alignment in Appendix C of the procedure which indicates SGE-V354 is to be in the open position.
Failure to comply with the procedure rendered ADV No.
178 inoperable from the control room when it was assumed that the valve in fact was operable.
This action is considered to be a violation of regulatory requirements (528/89-16-03).
9.
Im ro er Installation of Emer enc Diesel Generator Fuel Line Unit 1 62703 During the period, one of the equipment repair activities performed by the Unit 1 plant maintenance personnel was the repair of a leak in the Four Left (4L) fuel injection line at the fitting connection to the cylinder of the "A" emergency diesel.
Work package No.
351776 was developed to perform this work.
The work instruc tions specified that the installation of the line and the torquing of the nuts were to be performed in accordance with Technical Mahual
No.18-388,
"Palo Verde Nuclear Generating Station (PVNGS) Emergency Diesel Generator High Pressure Fuel Line Installation Procedure",
pages 2A and 2B, Steps 1 through 6.
These instructions apply to the installation of a new fuel line using new ferrules.
The work package also included a quality control (QC) sign-off confirming the installation was performed in accordance with the technical manual instructions.
Following the replacement of the line, a test run was conducted to confirm the leak was repaired.
Several minutes into the run the cylinder connection came loose causing fuel oil to spray.
An opera'tor in attendance immediately shut down the engine.
The following information was obtained by the inspector during his followup of the event.
o A review of the work package indicated both the installation work instruction line item and the QC confirmation line item had been signed off.
o A review of the technical manual showed that separate and specific instructions had been provided for the installation of a new ferrule/nut as well as the reinstallation of an old ferrule/nut.
In the case of the new ferrule/nut installation, the instruction stated that nut is to be torqued to 15 (plus-or-minus 2) foot-pounds and then an additional full turn.
In the case of the reused ferrule/nut installation, a torque of 15 (plus-or-minus 2) foot-pounds is required followed by an additional 1/4 (one-q'uarter)
turn.
o Based on discussions with the maintenance craftsman, the inspector learned that the craftsman had referred to the wrong instructions.
Mhile a
new ferrule/nut (initial installation)
was used, the instructions for the reinstallation of a previously used ferrule/nut assembly were implemented.
o Having used the wrong instructions, only a quarter turn of the nut was made, resulting in the ferrule/nut blowing off.
o The line was reinstalled properly and the engine tested satisfactorily.
The craftsman could not explain to the inspector his use of the wrong instruction except that it was a
human error..
He also stated that he had discussed the instruction with the QC inspector who had concurred in the procedure.
The inspector noted that while some additional clarification would improve the instruction, the instruction was adequately clear.
The licensee's corrective actions include the retraining of the crew involved in the repair work.
The Quality Control (QC) inspector will also be trained on this event, and the need to interpret work instructions correctly will be discussed with the QC staff.
In addition, separate generic maintenance procedures will be developed
for both the installation of new nut/ferrule assemblies and the reinstallation of used nut/ferrule assemblies.
The work package will carry only the applicable instruction.
These procedures will apply to all three units.
These actions should preclude the recurrences of this event.
The failure to follow approved procedures which resulted in the fuel line disconnecting from the cylinder is considered a violation of regulatory requirements (528/89-16-04).
10.
Multi le Electrical Load Sheds Units 1 and
71707 On March 20, 1989, a clearance was to be hung and the fuses pulled on the X-Winding synchronizing potential transformer of Unit 1 auxiliary transformer 1E-NAN-X02 to support maintenance on the 525 KV transmission lines.
The operator, having opened the electrical cabinet door, observed two drawers.
He noted the top drawer was labeled
"lE-NAN-S01 Bus PT" and also had a
permanent caution tag which read;
"Caution bus potential transformers are enclosed within - opening door will cause a
shed of this bus".
The lower drawer contained the label
"lE-NAN-X02 X-Winding'Sync PT".
The clearance document identified the component as
"lE-NAN-S01B Line PT, 1E-NAN-X02 X-Winding Sync PT".
The auxiliary operator selected the top drawer to be pulled but was unsure so he consulted with the control room reactor operator who concurred in his selection.
Upon pulling the drawer containing the fuses, power was disconnected to the lockout relays of the loads supplied by bus lE-NAN-SOl and the non-class loads were shed including two reactor coolant pumps.
b.
On April 12, 1989, during the isolation of power circuits to support. maintenance work on 525 KV insulators associated with the main transformer, electrical loads were inadvertently stripped from the Unit 2, 2E-NAN-S01 bus.
These loads were non-class but did include two reactor coolant pumps.
The inadvertent loss of the loads occurred when the operator erroneously opened the 1E-NAN-S01B bus bar potential transformer fuse drawer instead of the lE-NAN-X02, X-Winding synchronizing potential transformer fuse drawer.
This error was identical to the March 20, Unit 1 error.
The followup of these events by the inspector revealed the following:
1)
An inspection of the drawer labels disclosed that the
'equipment identification of the drawers at the three units were different.
In the case of the lower drawers which were to have been disconnected, the labels were as follows:
o 'nit.1 -
1E-NAN-X02 X-Winding Sync PT
o Unit 2 -
2E-NAN-S01A Circuit PT A-B, B-C, Fuse 0-54 o
Unit 3 -
3E-NAN-S01 Line PTs From Unit Aux Transformer 3E-NAN-XOl (Inspector's Note: This is a licensee typographic error, the designation should be 3E-NAN-X02).
2)
Both the Unit 1 and 2 clearances identified the drawers to be pulled as:
l(or 2)
E-NAN-S01B Line PT 1(or 2) E-NAN-X02 X-Winding Sync PT Unit 3 clearance identified the drawers as:
"3E-NAN-SOlA PT Potential XFMR For S01A (Line PT Fuses)".
3)
In the case of Unit 1, the auxiliary operator solicited help from the control room operator.
They both reached the conclusion that pulling what turned out to be the wrong drawer was a proper action.
In the case of Unit 2 the operator was confused by the difference in equipment identification in the clearance instruction and on the drawer label.
The Unit 2 auxiliary operator proceeded without a second consultation.
The Unit 1 event was being investigated as a Category 3 event.
Administrative Procedure 79AC-01P02 "Incident Investigation for Category 3 Incidents" does not provide a specific time for a report
'o be issued on a Category 3 event.
In addition, at the time Unit 1 did not feel that labeling was a problem at the other two units since it was not one at Unit 1.
The licensee's immediate plans to correct this problem included the use of consistent labels at the three units.
The label wording will be consistent to the design drawings and the work control clearance information will be made consistent with the drawings and labels.
A report of these experiences will be provided to the shift crews for training.
The need to evaluate other electrical equipment for consistent labeling was reviewed by management who committed to implement corrective actions as necessary.
The inspector concluded that these actions appeared to offer an appropriate solution to the problem.
c.
On April 4, 1989, while implementing a clearance to isolate normal service transformer 1E-NBN-X01, the auxiliary operator
'4 opened the electrical cubical to pull the drawer which would allow removal of the fuses.
The relays protruding from the back of the door were partially blocking access to the drawer and as the operator reached for the drawer, his key ring contacted two terminals on one of the under-voltage relays
'ausing the shed of non-class loads on bus 1E-NBN-SOl.
One of the factors contributing to this event was inadequate knowledge on the part of the auxiliary operator of the electrical shock hazards.
He believed the relay voltages were low and that there was no potential hazard to be close to the terminals.
Corrective actions will include personal training to the operator, implementation of a training lesson for non-licensed operators that emphasizes electrical safety hazards and an upgrade of the training material in the PVNGS Safety Manual for switching.
The relays which were contacted by the operators keys were found mounted differently in Units 2 and 3.
The relays in Unit 1 will be relocated as in Units 2 and 3.
While personnel error had been made, the equipment involved was not safety related and the plant did.not experience any operational control problems.
The inspector discussed these errors with utility management.
No violations or deviations of NRC requirements were identified.
ll.
Corroded Diesel Generator Intercooler Elbow - Unit 2 93702)
On April 12, 1989, an elbow fitting on the top of Unit 2 diesel generator (DG) "A" developed a leak of approximately 1 gallon per minute.
The leak was due to through wall corrosion of the steel elbow on the spray pond water side of the DG intercooler.
This appeared to be the same corrosion mechanism that caused previous failures of intercooler drain plugs on the Unit 3 and Unit 2 DGs.
On July 14, 1988, the Unit 3 "A" DG intercooler drain plug blew out when spray pond pumps were started.
A visual inspection of the plug showed signs of insufficient thread engagement due to corrosion of the plug.
The licensee's Engineering Evaluations department (EED)
initiated Engineering Evaluation Request (EER) 88-DG-058 in-response to the Unit 3 intercooler drain plug failure.
On July 17, 1988, EER 88-DG-059 was initiated to evaluate the different style'of drain plugs on the Unit 1 DGs.
On July 21, 1988, a second EER 88-DG-060. was initiated to evaluate a
Unit 1 "A" DG intercooler drain plug that was corroded excessively.
The interim resolution was to manufacture carbon steel plugs to replace the corroded plug.
EER 88-DG-064 was initiated on July 25, 1988, to evaluate Unit 1 "B" intercooler drain plugs that were
corroded and to combine the efforts of EERs 88-DG-58, 59, and 60.
This EER incorporated drain plugs with zinc anodes and allowed for carbon steel plugs to be used prior to manufacture of the zinc anode plugs.
On September 9, 1988; an intercooler drain plug failed on the Unit 2
"A" DG.
This failure was attributed to the same corrosion mechanism exhibited in the Unit 1 and
DG intercooler plugs.
A work order to inspect/replace Unit drain plugs had not been completed prior to this second event.
As discussed in the most recent SALP report, this was an example of weak problem identification since the same event had occurred on Unit 3 less than three months before.
EER-88-DG-064 was closed on November 7, 1988.
That EER stated that as of October 4, 1988, no work order had been initiated to install the new drain plugs in Units 1 and 3 and that the installation of the plugs should be raised to the highest priority.
Work orders initiated at Unit 2 were scheduled to be completed before October 5, 1988.
The EER also recommended establishing a Preventive Maintenance (PM) task to monitor the corrosion of the zinc anodes.
The initial frequency of the PM was suggested to be semi-annual.
The inspectors review of the April 12, 1989 intercooler elbow leak indicated that Work Order (WO) 00237201 was performed on September 15, 1987, to replace a similar elbow on Unit 2 "A" DG intercooler.
The WO indicated the elbow was removed in pieces but did not explicitly indicate corrosion was the cause of the damage to the elbow.
However, the WO indicated water was spraying from the elbow and the drawing and part number were the same as the April 12, 1989 failure.
The inspector made the following conclusions:
o Failure of the Unit 3 drain plug was not acted on aggressively to preclude a similar occurrence at Unit 2..
o The corrective action for the Unit 3 drain plug was not thorough in that it only addressed the specific problem of drain plugs and did not address other carbon steel components in the system susceptible to the same corrosion mechanism.
The matter was first discussed with the licensee at the time of the September 9, 1988, failure of the Unit 2 drain plug and in the most recent SALP report.
The'subject was again discussed with the licensee's management who acknowledged the licensee's comments and indicated agreement.
No violations or deviations of NRC requirements were identified.
Mid-Loo 0 erati ons - Unit 3 71707 The inspector observed mid-loop operation preparations, entry and exit in Unit 3.
The licensee's mid-loop activities, including
I.
h
problem resolution and responsiveness to NRC Generic Letter 88-17
, "Loss.,of Decay Heat Removal" were reviewed.
Finally, the effectiveness of the guality Audits and Monitoring (gA and M), and Independent Safety Engineering (ISE) oversight groups was assessed.
The inspector made the following observations:
Procedural adequacy.
Procedure 43A0-3ZZ22, "Loss of Shutdown Cooling (SDC)",
stated that if SDC flow-were totally lost while in Mode 5, operators should feed and bleed the steam generator secondary sides to provide for reactor coolant system (RCS) heat removal.
No recommended actions existed for the Mode 5 conditions when steam generators were unavailable due to mid-loop operations.
The NRC inspector identified this discrepancy and it was corrected by the li'censee prior to mid-loop operations.
2)
Procedure 430P-3ZZ16,
"RCS Drain Operations",
did not provide guidance for when or how to vent the SDC system.
Precursor indications such as abnormal flow noise or the appearance of air bubbles in the tygon tube level indicator were not addressed.
Specific valve numbers, and sequencing for venting operations were not addressed;
3)
Procedure 430P-3ZZ16,
"RCS Drain Operations",
as originally issued, contained an incorrect correction factor curve for the "A" RCS loop tygon tube level indication.
The incorrect curve assumed a different tygon tube connection point to the RCS than the one actually used.
Operators discovered the error during RCS drain operations when "A" and "B" loop levels became significantly different.
They stopped draining and corrected the error before proceeding.
However, the inspector noted that correction curve data supplied by engineering had been incorrectly incorporated into the procedure.
This is considered a violation of regulatory requirements (530/89-16-01).
4)
The surveillance test calibration procedure for the SDC flow meter, used to ensure Technical Specification minimum flow requirements, was found by the licensee to indicate approximately 160 gpm greater than actual flow due to the in-use fluid temperature of 90 degrees F being lower than the calibration temperature of 300 degrees F.
This instrument is an orifice flow restriction device with a differential pressure detector.
The inspector noted that this was a case of engineering data incorporated into a calibration procedure which resulted in an initially unrecognized actual difference between indicated and actual flow.
The licensee subsequently determined that due to conservatism of the minimum flow requirement, the indicated flow may be used without correctio Based on the above observations, the inspector concluded that; 1) procedures related to mid-loop operations were in some cases incomplete and inaccurate, and 2) there appeared to be a lack of control over the inclusion of engineering supplied data into operations and instrument calibration procedures.
Licensee management committed to a reassessment of the mid-loop operations procedures, including loss of SDC, with the objective of reverifying Generic Letter 88-17'equirements, ensuring the adequacy of engineering input, and incorporating all lessons learned from Unit 3, and completing the necessary revisions and training prior to any further mid-loop operations with fuel in the vessel (530/89-16-02).
Second, licensee management committed to reviewing the policies and controls associated with the'xchange and review of information between the engineering and standards organizations.
This item will be followed up in a future inspection (530/89-16-04).
b.
Operations during mid-loop condition.
1)
Following entry into mid-loop operation, operators attributed the appearance of "growling" and "rumbling" flow noises, emanating from specific locations in the SDC flow path, to be caused by normal flow dynamics.
Consideration of possible air entrainment was apparently not made, even though the noises appeared only after the plant was placed in a mid-loop condition.
Operators were aware of the noises for approximately two days prior to notifying a system engineer.
2)
On March 26, 1989, operators attempted to minimize or eliminate the flow noise by slightly adjusting various throttle valves.
In doing so, they increased SDC flow from 4100 gpm, the maximum flow recommended by procedure, to 4250 gpm.
The procedure indicated that the 4100 gpm recommendation was based on preventing vortexing or air entrainment in the SDC flow path.
3)
On March 27, 1989, subsequent to increasing SDC flow to 4250 gpm, air bubbles appeared in the tygon tube level indicator associated with the operating SDC train.
Operators reduced SDC flow and eliminated the air bubbles.
A system engineer walked down the flow path, but made no immediate recommendations.
4)
On March 28, 1989, one day later, air bubbles reappeared in the same tygon tube indicator,,and the system engineer concurred with operations that the SDC system should be vented.
An estimated 100 gallon volume of air was then vented from the system.
Based on the above observations, the inspector concluded that the plant experienced vortexing and air entrainment during mid-loop operations.
This is considered a violation of regulatory requirements (530/89-16-03).
In addition, operators appeared to
inappropriately attempt to reduce flow noise by exceeding procedural recommendations to limit total SDC flow.
Finally, the onsite engineering staff was slow to recommend corrective action.
Licensee management restated their commitment to ensuring that all appropriate operations and engineering staff, including management, are briefed on the significance of these events prior to the next, mid-loop operation with fuel in the vessel.
Furthermore, licensee management commited to establishing, by adequate technical means, the actual margin to vortexing prior to the next mid-loop operation.
This is part of open item (530/89-16-02),
addressed earlier in the section.
In addition, the inspector noted that the licensee was pursuing a
change to the minimum SDC flow required by Technical Specifications.
c.
Evaluation of Oversight Group Effectiveness.
The inspector reviewed gA Monitor Report No.
MOR89-0025 and Independent Safety Engineering (ISE) surveillance report No.89-012, both covering Unit 3 mid-loop operations.
The inspector assessed the degree to which these reports formed a self critical review of the Unit 3 mid-loop operation, and their emphasis on corrective actions needed prior to another unit entering a mid-loop condition.
The inspector determined that neither report recommended any corrective action to be completed prior to the next mid-loop operation.
The gA report was critical only of some differences between the training lecture given to the Technical Staff and the final approved RCS Drain Operation procedure.
The ISE report, under
"Recommendations and Future Actions", only committed the ISE group to evaluate the inaccuracy of the SDC flow instrument'nd to review changes to the licensee's commitment to monitor the tygon tube level indications.
Neither the gA or the ISE reports were critical of the adequacy of procedures in use.
The inspector concluded that the gA and ISE critiques were
~ineffective in recognizing the scope and depth of needed changes to procedures, organizational interfaces, and operating policy.
Licensee management acknowledged these concerns and stated that renewed emphasis would be given for these groups to provide more critical reviews.
In conclusion, the licensee's preparations and conduct of mid-loop operations, following their commitments to NRC Generic Letter 88-17, did not prevent several problems from occurring, including entry of the plant into a vortexing condition which is a precursor to air binding a SDC pump and loss of SDC flow.
The licensee's corrective action in response to these concerns will be carefully reviewed.
No violations or deviations of NRC requirements were identifie I
Review of Licensee Event Re orts - Units 1 2 and
90712 92700 The following LERs were reviewed by the Resident Inspectors.
Based on the information provided in the report, it was concluded that reporting requirements had been met, root causes had been identified, and corrective actions were appropriate.
The below listed LERs are considered closed.
Unit 1 LER NUMBER DESCRIPTION 88-02-LO Nonconservative Set oints on Hi Lo Power Tri
~Closed This licensee event report is discussed in NRC Inspection Report 50-528/88-41.
This item remained open pending the issuance of a supplemental report.
The supplemental report has been issued, reviewed, and appeared adequate.
This item is closed.
88-20-LO Control Element Assembl CEA Sli a e.
Closed)
This report discussed the slippage of CEAs No.
64 and No.
57 during the performance of a monthly CEA operability check on December 10, 1988.
This event is discussed in NRC Inspection Report 50-528/88-41.
At the time the inspector concluded the licensee's actions to recover from the condition were appropriate so that for this report this item is considered closed.
Subsequent evaluation of this event by the licensee and Combustion Engineering has resulted in a procedure revision which requires a plant trip whenever there is the simultaneous slippage of two CEAs.
This item is closed.
88-26-LO Control Element Assembl CEA Sli a
e Due to This report discussed the slippage of CEA No.
64 during the performance of a monthly CEA operability check on November 5,
1988.
The cause for the slippage had been determined to be a gr'ound on the coil of the lower gripper assembly.
The correction of the problem required the plant to be in Mode 6 so that the permanent corrective action will be accomplished during the current refueling outage.
The immediate corrective actions required by technical specifications were completed by the licensee.
At the time of the slippage the inspector's review of those actions concluded the licensee responded properly to the CEA slippage.
A Technical Specification change was also approved defer ring the testing of CEA No.
64 until after the refueling outage.
This item is closed.
89-01-LO Missed Reactor Coolant Pum RCP Vibration Monitorin and En ineerin Anal sis.
Closed This event involved the failure to perform a daily engineering analysis of the Unit 1 reactor coolant pump vibrati on data.
This
"failure was due to a personnel error.
The individual responsible was instructed on the need for timely performance of routine tasks.
Periodic reviews by the inspector of the timeliness of the engineering analyses did not reveal any similar type violations.
This item is closed.
Unit 3 LER NUMBER DESCRIPTION 89-04-LO Closed Emer enc Diesel Generator Rocker Arm Failure/
89-04-L1 0 en En ineered Safet Features ESF Actuation The reports discussed the failure of an exhaust valve rocker arm on the 3A Emergency Diesel Generator (EDG) and a spurious Main Steam Isolation Signal (MSIS) actuation during the subsequent plant cooldown to Mode 5.
The rocker arm failure was reported in Inspection Report 50-530/88-41.
The licensee's corrective actions were evaluated as adequate in that report.
In addition, the failure mechanism which caused the spurious MSIS was described in LER 89-004-Ll.
This mechanism was a high resistance condition across relay contacts that are required to open to perform -their safety function.
Although the event in question occurred because of insufficient cur rent flow due to the abnormally high contact r'esistance, the LER had not addressed the root cause of what mechanism was creating the high resistance condition and whether that mechanism could prevent the contacts from opening, thereby preventing fulfillment of its safety function.
The licensee agreed to respond to this concern.
This LER will remain open pending resolution of this concern.
The inspector considered all other actions to be adequate.
Review of Periodic and S ecial Re orts - Units
2 and
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
'ontained the information required to be reported by NRC requi rements; test results and/or supporting information were consistent with design predictions and performance specifications; and the validity of the reported information.
Mithin the scope of the above, the following reports were reviewed by the inspector.
Unit 1 I
o Monthly Operating Report for March, 1989.
Unit 2 o
Monthly Operating Report for March, 198 Unit 3 o
Monthly Operating Report for March, 1989.
No violations of NRC requirements or deviations were identified.
The inspector met with licensee management representatives periodically during the inspection and held an exit meeting on April 26, 1989.
During the exit meeting, the inspector emphasized the following:
o Inadequate control of mid-loop operations at Unit 3 from an operations, engineering and oversight perspective.
o The failure to control surveillance test procedure acceptance criteria.
o The failure to maintain control of valve status rendering the only operable ADV inoperable at Unit l.
o The failure to communicate lessons learned from one unit to the other units in the case of electrical load sheds at Units 1 and 2.
o The failure of maintenance and gC personnel to perform the correct work order steps in replacing a diesel fuel line at Unit l.
o The failure to generically resolve the root cause of failures in the case of the Unit 2 intercooler drain plug corrosion.
The licensee acknowledged the inspectors'oncerns and acknowledged the commitments that the inspectors noted in the other sections of this report.