IR 05000397/1991018
| ML17286A966 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/18/1991 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17286A959 | List: |
| References | |
| 50-397-91-18, NUDOCS 9108050065 | |
| Download: ML17286A966 (20) | |
Text
Report No:
Docket No:
Licensee:
Facility Name:
Inspection at:
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
50-397/91-18 50-397-Washington Public Power Supply System P. 0.
Box 968 Richland, WA 99352 Washington Nuclear Project No.
(WNP-2)
WNP-2 site near Richland, Washington Inspection Conducted:
May 13 - June 23, 1991 Inspectors:
R.
C. Sorensen, Senior Resident Inspector D. L. Proulx, Resident Inspector Approved by:
Summary:
P.
H.,
hnson, Chief Reacto Projects Section
Date Signed Ins ection on Ma 13 - June
1991 Ins ection Re ort No. 50-397 91-18)
'reas Ins ected:
Routine inspection by the resident inspectors of control room operations, operational safety verification, surveillance program, main-tenance program, licensee event reports, special inspection topics, procedural adherence, and review of periodic reports.
During this inspection, Inspection Procedures 61726, 62703, 71707, 90712, and 92700 were utilized.
I I
Safet Issues Mana ement S stem SIMS Items:
None.
Results:
General Conclusions and S ecific Findin s
Si nificant Safet Matters:
None.
Summar of Violations and Deviations:
Two violations were identified.
One involved a fire protection program violation and the other involved several clearance orders which had not received the required independent verifications.
0 en Items Summar
Seventeen LERs were closed; two new followup items were opened.
9108050065 9507iB PDR
. ADOCK 05000392 Q
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DETAILS 1.
Persons Contacted
.
- J. 'Baker, Plant Manager
- L. Harrold, Assistant Plant Manager C.
Edwards, guality Control Manager
- R. 'l<ebring, Plant Technical Manager'.
Graybeal, Health Physics and Chemistry Manager
'.
Harmon, Maintenance Manager A. Hosier, Licensing Manager
- S. Davison, guality Assurance Manager R. Koenigs, Generation Engineering Manager
- S. McKay, Operations Manager
- J. Peters, Administrative Manager G. Gelhaus, Assistant Technical Manager W. Shaeffer, Assistant Operations'anager The inspectors also interviewed various control room operators, shift supervisors and shift managers, maintenance, engineering, quality assurance, and management personnel.
- Attended the Exit Meeting on June 25, 1991.
2.
Plant Status At the start of the inspection period, the plant was in Node 5.
Mode 4 was entered on Hay 28, where the plant remained until the end of the inspection period..
3.
0 erational Safet Verification 71707 a.
Plant Tours The following plant areas were toured by the inspectors during the course of the inspection:
b.
The Reactor Building Control Room Diesel Generator Building Radwaste Building Service Mater Buildings Technical Support Center Turbine Generator Building Yard Area and Perimeter following items were observed during the tours:
(1)
0 eratin Lo s and Records.
Records were reviewed against Technical Specification and administrative control procedure require'ments,
-2-Monitorin Instrumentation.
Process instruments were observed for correlation between channels and for conformance with Technical Specification requirements.
~Etif'
i
.
C t
d hift kg h
d for conformance with 10 CFR 50.54.(k), Technical Specifica-tions, and administrative procedures.
The attentiveness of the operators was observed in the execution of their duties and the control room was observed to be free of distractions such as non-work related radios and reading materials.
E uiment Lineu s.
Valves and electrical breakers were veri-fied to be in the position or condition required by Technical Specifications and administrative procedures for the applicable plant mode.
This verification included routine control board indication reviews and conduct of partial system lineups.
Techni.cal Specification limiting conditions for operation were verified by direct observation.
E ui ment Ta in
.
Selected equipment, for which tagging requests had been initiated, was observed to verify that tags were in place and that the equipment was in the condition specified.
On Hay 24, 1991, the inspector reviewed selected clearance orders from the caution and danger tagout logs to verify that equipment was adequately isolated for work and that the associated administrative requirements were properly followed.
The inspector noted that PPH 1.3.8C,
"Danger Tag Clearance Order (Hanual)", Revision 0, section 4.2. 18, states,
"Independent Verification is required when the component is SAFETY RELATED (gC 1)...".
PPH 1.3.8A, "Caution Tag Order,"
Revision 0, Section 6.B.2.c.(6),
states in part, "Ifwork is associated with Safety-Related...
Systems, an independent verification of component position is required by a second qualified individual."
However, the inspector noted that the following clearance orders on safety related systems had not received the required independent (second) verification:
Clearance Order No..
91-6-C005 (Caution)
91-5-C094 (Caution)
91-5-C004 (Caution)
90-9-C021 (Caution)
90-12-C084 (Danger)
~Sstem Containment Atmospheric Control t
~
Reactor Core Isolation Cooling Inverter No.
High Pressure Core Spray Leak Detection System This is an apparent violation of Technical Specifications, Section 6.8. 1 (Violation 397/91-18-01).
The inspector brought these d'iscrepancies to the attention of the Shift Hanager, and
-3-(6)
the required second verification was completed for each of these clearance orders on the subsequent shift; General Plant E ui ment Conditions.
Plant equipment was observed for indications of system leakage, improper lubrica-tion, or other conditions that could prevent the system from fulfillingits functional requirements.
Annunciators were observed to ascertain their status and operability.
On June 7, 1991 the inspector noted that gauge 2-SW-FIS-9 was indicating 325 gallons per minute (gpm) with no flow through the system.
This gauge is used to set the flow rate for Standby Service Water (SW) cooling to the High Pressure Core Spray (MPCS, or Division 3)
Emergency Diesel Generator (EDG).
The inspector was concerned that since the required SW flow rate to this EDG is 910 +/- 45 gpm, a 325 gpm error in the nonconservative direction could result in inadequate cooling of the EDG.
The inspector informed the system engineer (who was unaware of this condition)
and the instrumentation and controls supervisor.
The licensee informed the inspector that the gauge in question was not due for recalibration until April of l992.
However, a
calibration check was performed and the. gauge failed:
The
,
licensee initiated a Haintenance Work Request (HWR) to troubleshoot and repair the gauge, and found a failed micro-switch within the gauge, which was replaced.
However, a
subsequent calibration check also failed, so the licensee replaced the gauge with a new one which was successfully calibrated.
The flow balance was reperformed for the HPCS SW system, and the. system appeared to be operating properly.
The inspector requested that the licensee evaluate. whether or not the calibration frequency should be increased for 2-SW-FIS-9,
,and the System Engineer stated that this was an isolated occurrence and no increase in calibration frequency was warranted.
The plant manager, agreed that this type of deficiency should have been identified by the System Engineer or by Operations personnel during tours rather than being brought to their attention by the NRC.
Fire Protection.
Firefighting equipment and controls were observed for conformance with administrative procedures.
f
~
On Hay 13, 1991 at approximately 8:00 a.m.,
the inspector observed that fire door R402 at the northeast corner of the 522-foot elevation of the reactor building was propped open, with no Fire Protection System Impairment Checklist evident.
Further inquiry revealed that one had not been completed.
Since the fire door was propped open to support work on a
Reactor Closed Cooling (RCC) heat exchanger on the 548'levation of the reactor building, an Impairment Checklist was required by PPM 1.3. 10, Fire Protection Program.
This failure to complete a Fire Protection System Impairment Checklist is an apparent violation of Technical Specifications, Section
l 1,
,f t
II',
t
6.8. l.g (Violation 397/91-18-02).
The inspector brought the open fire door to the attention of the 'Fire Marshall, who acknowledged the inspector's finding.
Later that same day.the Fire Marshall identified a number of other fire doors that had been propped open.
The inspector subsequently brought this issue to the attention of plant management.
The Fire Marshall issued a memorandum to applicable plant staff informing them of several problems identified in the area of fire protectio'n, and'discussing expectations for complying with fire protection requirements at WNP-2.
(8)
Plant Chemistr
.
Chemical analysesand trend results were reviewed for conformance with Technical Specifications and administrative control procedures.
(9)
Radiation Protection Cont'rois.
The inspectors periodically observed radiological protection practices to determine whether the licensee's program was being implemented in conformance with facility policies and procedures and in compliance with regulatory requirements.
The inspectors also observed compliance with Radiation Work Permits, proper wear-ing of protective equipment and personnel monitoring devices, and personnel frisking practices.
Radiation monitoring equipment was frequently monitored to verify operability and adherence to calibration frequency.
(10) Plant Housekee in
.
Plant conditions and material/equipment storage were observed to determine the general state of clean-liness and housekeeping.
Housekeeping in the radiologically controlled area was evaluated with respect to controlling the spread of surface and airborne contamination.
(11) ~Securit
.
The inspectors periodically observed security practices to ascertain that the licensee's implementation of the security plan was in accordance with site procedures, that the search equipment at the access control points was opera-tional, that the vital area portals were kept locked and alarmed, and that personnel allowed access to the protected area were badged and monitored and the monitoring equipment was functional.
En ineered Safet Feature Walkdown Selected engineered safety features (and systems important to safety)
were walked down by the inspectors to confirm that the systems were aligned in accordance with plant procedures.
During the walkdown of the systems, items such as hangers, supports, electrical power supplies, cabinets, and cables were inspected to determine that they were operable and in a condition to perform their required functions.
Proper lubrication and cooling of major components were observed for adequacy.
The inspectors also.
verified that certain system valves were in the required position by both local and remote position indication, as applicabl '
-5-Accessible portions of the following systems were walked down on the indicated dates.
~Sstem Diesel'Generator Systems, Divisions 1, 2, and 3.
Low Pressure Coolant Injection (LPCI)
Trains "A", "B", and
"C"
'ow Pressure Core Spray (LPCS)
High Pressure Core Spray (HPCS)
Reactor Core Isolation Cooling (RCIC)
Standby Liquid Control (SLC) System Standby Service Water System 125V DC Electrical Distribution, Divisions 1 and
250V DC Electrical Distribution Dates June 7,
June 7,
June
June
'June
June
June
June
June
Ho violations or deviations were identified.
4.
Surveillance Testin 61726 a
~
Surveillance tests required to be performed by the Technical Specifications (TS) were reviewed on a sampling basis to verify that:
(1)
a technically adequate procedure existed for performance of the surveillance tests; (2) the surveillance tests had been performed at the frequency specified in the TS and in accordance with the TS surveillance requirements; and (3) test results satisfied acceptance criteria or were properly dispositioned.
b.
Portions of the following surveillance tests were observed by the inspectors on the dates shown:
d a
~
7.4,3.4. 1.2 ATWS-RPT-ARI RPV Press Hi Test 7.4.1.3.5.3 Scram Accumulator Check Valve Operability Test 7.4.5. 1. 19 HPCS Suction Transfer Test Dates Performed June
June
June
C.
The inspector witnessed the preparation for, and portions of the conduct of, the Primary Containment Integrated Leak Rate Test (ILRT),
PPH 7.4,6. 1.2. 1, on June 5, 6, and 7.
The inspector verified selected prerequisites and witnessed pressurization of the
-6-primary containment to about 17 psig.
Pressurization was halted prior to reaching the test pressure of approximately 35 psig due to a crack in the containment purge piping outside of the containment boundary.
The containment purge path into the drywell had been selected as the primary pressurization path for the leak rate test.
The existence of the crack precluded continued pressurization by that path so the alternate path was employed utilizing two spare containment penetrations.
Test pressure was eventually achieved on the afternoon of June 7, but suppression pool level was decreasing slowly, indicating suppression pool leakage, which would affect the-validity of the test results.
Attempts were made to determine the location of the leakage.
RHR "C" was finally determined to be the location because the leakage stopped when the suction valve RHR-V-4C was close'd.
However, the destin'ation of the leakage had not been firmly established by the end of the inspection period.
Preliminary indications were that RHR "C" had a leakage path that led outside of secondary containment.
The ILRT was successfully completed on June 9 with RHR-V-4C in the closed position.
Leaka'ge was measured to be approximately 0:21 weight percent per day, well within the
.375 weight. percent per day, acceptance criteria.
RHR-V-4C is a containment isolation valve that passed its individual local leak rate test.
The inspector determined through consultations with NRR that conduct of the ILRT'ith RHR-V-4C closed was valid.
However,= the Technical Specifica-.
tions require a program, which the licensee has established and implemented, to monitor certain systems for potential post-accident leakage.
The RHR system is included in this'rogram and it appeared to the inspector that the appropriate action would be to find the leaking components and correct the problem.
During the exit meeting, Plant Management committed to take steps, prior to plant restart, to locate and evaluate the source of the leakage.
(Followup Item 397/91-18-03)
No violations or deviations were identified.
I 5.
Plant Maintenance 62703 During the inspection period, the inspector observed and reviewed
'ocumentation associated with maintenance and problem investigation activities to verify compliance with regulatory requirements and with administrative and maintenance procedures, required QA/QC invol.vement, proper use of clearance tags, proper equipment alignment and use of jumpers, personnel qualifications, and proper retesting.
The inspector verified that reportability for these activities'was correct.
The inspector witnessed portions of the following maintenance activities:
Descri tion Date Performed Repair Woodward Governor on Division II EDG per AR 4225 May 29
f I
t
-7-Replace RPS "B" HG per AR 4461 Install New Recirculation Pump Vibration Monitoring System No violations or deviations were identified.
6.
Licensee Event Re ort LER Followu 90712 92700 June
June
a.
The following LERs associated with operating events were reviewed by the inspector.
Based on the information provided in the reports it was concluded that reporting requirements had been met, root causes had been identified, and corrective actions were taken or initiated as appropriate.
The below LERs are considered closed.
LER NUMBER DESCRIPTION 90-14 90-15 HPCS Pump Suction Valve Switchover on High Suppression Pool Level Due to Procedural Inadequacy Diesel Fuel Oil Analysis Verification Time Requirement Exceeded
- Procedure and Personnel Error 90-20 91-01 91-02 91-03 91-07 91-09 91-11 90-25-01 91-12 Overheating of EDG Control Cabinet Components Could Cause Failure Due to Inadequate Design RCIC-V-8 ESF Actuation Due to Failed Component in Leakage Detection System Jet Pump Operability Testing Not in Compliance Mith Technical Specification Requirements Surveillance Testing of Standby Gas Treatment System Not in Compliance With Technical-Specifications Requirements Reactor Scram and Shutdown Cooling Isolation While Conducting Excess Flow Check Valve Testing Reactor Water Cleanup (RWCU) System Isolation-Caused by Recorder Wiring Error Shutdown Cooling Isolation Due to Inadvertent Mire Cutting During Recorder Replacement Inoperability of HPCS Caused By Improper Torque Switch Setting on HPCS-V-23 Nanual Scram in Node 5 on Low Scram Air Header Pressure Caused by Temporary Air Hose Failure
0 h
-8-91-14 Shutdown Cooling Isolation Due to Inadequate Design Drawing Closed LERs 90-18 90-22 90-22-01 90-23 and 90-23-01
-- Various En ineered Safet Feature Isolations of the Containment Instrument Ai ffC1 S <<
These LERs involved the ESF isolation of the nitrogen portion of the CIA system from the ADS nitrogen supply.
Certain aspects of these LERs demonstrated weakness in documenting all root causes and corrective actions.
This was especially true of 90-22, in which the cryogenic tank was depleted to the point of depressurizing the CIA system during primary containment inerting.
The plant manager had agreed to submit a more complete supplemental LER if appropriate.
The licensee submitted LER 90-22-1 on April 22 which was much more comprehensive in dealing with root causes and corrective actions.
Further, the inspector verified a number of the licensee's correc-tive actions, including procedure enhancements for nitrogen inerting of containment, and instrument setpoint changes for the cryogenic tank low level alarm.
The inspector concluded that the licensee's response and corrective
'ctions were appropriate.
These LERs are considered closed.
0 en LER 91-05 - Ox en Concentrat on in Su ression Chamber Not Verified Per Technical S ecification Re uirements This LER involved the failure to have in place a procedure to ensure appropriate verifications of oxygen concentration in the wetwell.'he problem was discovered as a result of a surveillance procedure which tests the wetwell spray mode of RHR.
Wetwell spray causes cooling of the wetwell airspace, resulting in a slightly negative pressure with respect to the reactor building.
This causes a subsequent opening of the, reactor building-to-wetwell vacuum breakers, drawing in air (and oxygen) to the wetwell.
This occurred on triarch 29.
On Harch 30, the Shift Manager decided to sample the wetwell for oxygen concentration, and it was determined to be 3.9%, greater than the Technical Specification limit of 3.5%.
The applicable action statement was entered and action was taken to restore oxygen concentration to within the limit. It was later determined that no mechanism was in place to ensure monitoring of wetwell oxygen concentration every. seven days, as required by the Technical Specifications, or following plant evolutions which could adversely affect the oxygen concentration.
The licensee took immediate action to revise PPH 7.0.0 to include daily verification of wetwell oxygen concentration.
However, the inspector disagreed with the licensee's conclusion as to the minimal safety significance of this issue.
Control room operators indicated that the reactor building-to-wetwell vacuum breakers have opened during previous testing of suppression pool spray.
In addition, this surveillance test is performed quarterly.
These
(
i l
-9-facts, coupled with the fact that no mechanism was in place to monitor wetwell oxygen concentration every seven days, make it likely that oxygen concentration had exceeded the Technical Speci-fication limit on a number of occasions in the past, for long periods of time, perhaps to a value greater than 3.9%.
It is therefore possible that, had a design basis event occurred, it could have occurred with wetwell oxygen concentration greater than 3.5%.
The licensee agreed to reassess their conclusion of safety significance and analyze the potential effects on containment with an initial oxygen concentration of 3.9%.
This LER remains open.
d.
Closed LER 91-06 - Plant Shutdown Re uired b
Technical S ecifications Due to Ino erable ED As discussed in this LER, the licensee declared the Division
EDG inoperable and initiated a plant shutdown to begin the refueling outage because of foreign material found in the generator bearing lube oil sump.
Subsequent investigation by the licensee determined that the debris observed in the sump had actually resulted from the bearing failure.in May 1990, but had not been completely removed as part of the repair process conducted at that time.
The licensee's root cause assessment concluded that the complex geometry and orientation of the lube oil sump do not facilitate effective cleaning of the sump, a consideration which was not well understood in 1990.
As a result of the lessons learned from this experience, the licensee performed more extensive cleaning of the sump in April 1991, including steam cleaning in both the normal and inverted orientations.
In addition, a magnet was used to collect bearing fragments from the sump, in that all bearing components other than the bearing races are made of magnetic materials.
The licensee concluded that this improved cleaning method removed all but trace amounts of debri,s from the sump.'earing oil samples after post-maintenance testing indicated levels of foreign material which were substantially below those observed before the refueling outage.
This LER is closed.
No violations or deviations were identified.
7, Review of Periodic and S ecial Re orts 90713 Periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9. 1 and 6.9.2 were reviewed by the inspector.
This review included the following considerations:
the report contained the information required to be reported by NRC requirements, and the reported information appeared valid.
Within the scope of the above, the following report was reviewed by the inspector.
o Monthly Operating Report for April, 1991.
No violations or deviations were identifie t fh
t
-10-The inspectors met with licensee management representatives periodically during the report period to discuss inspection status, and an exit meeting was conducted with the indicated personnel (refer to paragraph 1)
on June 25, 1991.
The scope of the inspection and the inspectors'indings, as noted in this report, were discussed with and acknowledged by the licensee representatives.
The licensee did not identify as proprietary any of the information reviewed by or discussed with the inspector during the inspectio ~
'