IR 05000382/1991003
| ML20217B400 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/28/1991 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20217B373 | List: |
| References | |
| 50-382-91-03, 50-382-91-3, NUDOCS 9103120099 | |
| Download: ML20217B400 (13) | |
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V.S. NUCLEAR REGULATORY C0iHISSION
REGION IV
NRC Inspection Report No: 50-3B2/91-03 Docket No: 50-382 License No: NPF-38 Licensee:
Entergy Operations, Incorporated P.O. Box B-Killona, Louisiana 70066 Facility Name: Waterford Steam Electric Station, Unit 3 (Waterford 3)
Inspection At: Taf t, Louisiana Inspection Conoucted: January S through February 19, 1991 Inspectors:
W. F. Smith, Senior Resident Inspector Project Section A, Division of Reactor Projects S. D. Butler, Resident Inspector Project Section A Division of Reactor Projects E
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Approved:
1. t. Westernian, Lntet, Froject dection A Uate Inspection Summary Inspection Conductpd January 9 through February 19, 1991 (Report 50-382/91-03)
Areas Inspected:
Routine, unannounced inspection of plant status, onsite followup of events, monthly maintenance observation, monthly surveillance observation, operational safety verification, followup of previously identified items, licensee event report followup, and preparation for refueling.
Results: A noncited violation was identified in paragraph 3.1 involving an inadequate procedure for leak testing a containment penetration. This illustrated a weakness in the engineering (development and technical reviews of test procedures in the inservice testing IST) area. As a result, the n'aximum allowed test pressure was exceeded; however, subsequent evaluation determined that no system damage occurred. The licensee self-identified, documented, and corrected the problem in a satisfactory manner, thus a violation was not cited.
The licensee's handling of a potential single failure in the steam bypass control system (SBCS), brought to their attention by an event at Palo Verde Unit 3, appeared corrmensurate to the safety significance of the item. However, these appeared to be a long delay from the time the problem was identified in late October 1990 until compensatory action was taken in late February 1991.
9103120099 910304 PDR ADOCK 05000382 O
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A violation was identified in paragraph 4.1 involving a case where a postmodification test procedure was not reviewed by the pknt operations review comittee (PORC) and approved by the general manager as required by Technical Specifications. This violation was cited because a similar violation occurred
.in June 1990, and the licensee's corrective actions did not prevent a recu rrence.
In addition, the licensee was not able to identify what additional corrective actions will be taken to correct the problem and prevent further violations. Also in paragraph 4.1, a noncited violation was identified involving an inadequate design change work package.
Although the licensee took prompt corrective action and documented the problem in their corrective action program, this was indicative of a weakness in engineering and technical support.
The licensee has continued to show strengths in refueling outage preparations and planning. There is good managenent support and involve,aent.
It appeared that the licensee was more successful than in the past in completing the it.ajority of design change packages prior to the start of the outage. New fuel receipt handling and inspection activities were found to be well executed and completed in accordance Lith approved procedures.
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DETAILS
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1.
PERSONS CONTACTED'
1.1 Principal Licensee Employees
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J. R. McGaha, General Manager, Plant Operations
- P. V. Prasankumar, Technical Services Manager D. F. Packer, Operations and Maintenance Manager
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- A. S. Lockhart, Quality Assurance Manager
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D. E. Baker, Director, Operations Support and Assessments
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R. G.. Azzarello, Director, Engineering
- T. P. Brennan, Design Engineering Manage:r W. T. Labonte, Radiation Protection Superintendent
- P. M. Helancon, Reactor Engineering Supervisor
- G. M. Davis, Events Analysis Reporting & Response Manager-
- K.= T. Walsh, Events Analysis & Reporting Supervisor R. F. Burski, Director, Nuclear Safety
- P. A. Gropp, System Engineering Supervisor-Mechanical.
L. W. -Laughlin, Licensing Manager
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1*T. J. Gaudet, Operational Licensing Supervisor
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i 4*J. G. Hoffpauir, Maintenance Super ntendent -
'*R. S. Starkey, Operations Superintendent A. G. Larsen, Assistant Maintenance Superintendent, Electrical D. T. Dormady,. Assistant Maintenance Superintendent, Mechanical D. C. Matheny, Assistant Maintenance' Superintendent, Instrumentation
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and Controls-
'*L. L.; Bass, Engineering Management Trainee -
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- Present at exit-interview.
In addition to the above personnel, the inspectors held discussions with
=various operations, engineering, technical support, maintenance, and
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administrative members of the licensee's staff.
2.
PLANT-STATUS--(71707)
,The plant was operated at full power from the beginning of this inspection-period on January 9 through February 19, 1991, except for'a few hours on
, January 18,'1991, when power was reduced to'approximately 90 percent-for 1 routine turbine valve testing.
In addition,Lfrom February 4-9, 1991, power
'was reduced to 65 percent, as requested by the system dispatcher, to
- accomodate reduced loads on the' power grid. - Planning and preparations have been underway to support the fourth refueling outage, which has been scheduled to-comence on March 15,.1991.-
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3.
DNSITE FOLLOWUP OF EVENTS (93702L i'
l 3.1 Overpressurization of Containment Penetration During Surveillance Test
On January 8,1991, the licensee entered Nuclear Operations Procedure N0P-19,
"Nonconformance/ Indeterminate Qualification Process," after Containment Penetration No.10 (containment purge inlet) was inadvertently overpressurized during performance of local leak rate testing in accordance with Surveillance Procedure OP-903-112, Revision 2, " Containment Purge Valve Leak Test." The penetration was pressurized to 51.5 psig for approximately 2 minutes before i
the individual performing the test recognized the condition and reduced pressure to 44-46 psig, which was the pressure specified for the test. Once the l
shift supervisor was notified, the test was continued and completed with satisfactory results. No visible damage was found on the penetration, but the licensee initiated an engineering evaluation to verify that the penetration had not been overstressed.
The inspectors reviewed the condition identification (CI273081) engineering evaluation, completed January 9, which concluded that the penetration and isolation valves, rated at 50 and 75 psig, respectively, were not overstressed by the event. The evaluation also concluded that the cause of the event was an inadequate procedure which required air flow to the penetration to be controlled with a valve that was downstream of the test gauge. The procedure was changed prior to performance of the test on Peretration No. 11 (containment purge outlet) on January 10.
This test was witnessed by the inspector and no problems were noted with the revised procedure. The inspectors expressed concern that Procedure OP-903-112 had been through the licensee's procedure upgrade program in June 1990 and was revised again in December 1990 to provide the deficient test apparatus.
The licensee
. acknowledged the problem and, as part of their corrective action, comitted to revise the other two affected procedures (0P-903-113 and OP-903-114) prior to use and no later than March 15, 1991. Other corrective actions included one-on-one discussions between licensee management and the personnel involved
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in tests and test procedure development to review lessons learned from problems such as those identified. Performing a surveillance test using an inadequate procedure is a violation of NRC regulations. However, this licensee-identified violation is not being cited because the criteria specified in Section V.G of the NRC's Enforcement Policy was satisfied. The inspectors will follow up to verify satisfactory completion of the licensee's corrective actions under Inspector Followup' Item (IFI) 382/9103-01.
3.1.1 Conclusions The licensee's actions to self-identify, document, and initiate corrective action when the penetration overpressurization event occurred was timely and appropriate, which shows a strength in the licensee's corrective action program. The licensee promptly informed the resident inspectors when the problem arose. However, the incorporation of a deficient test apparatus and inadequate procedural method, followed by acceptance of the deficiency during the technical reviews, was an indication of weakness in the technical support and review process. The inspector discussed this issue with the license.
3.2 Design Basis Uncertainty on Steam Bypass Control System (SBCS)
On October 20, 1990, the Palo Verde Unit 3 reactor tripped f rom full power.
The core prntection calculators initiated the trip af ter the departure from nucleate boiling ratio exceeded the trip setpoint.
The trip setpoint was exceeded when a-diode failed in.the turbine SBCS, opening all seven of the in-service steam bypass valves (see LER 530/90-007 and NRC Inspection Report 50-530/90-45). The Palo Verde FSAR states that only one bypass valve would open upon certain SBCS failures. Since the Waterford 3 FSAR similarly states that no single equipment failure nor operator error would cause the unwanted opening of more than one bypass valve, Region IV staff alerted the licensee at Waterford 3 to the event at Palo Verde.
In addition, Combustion Engineering (CE) issued
"Infobulletin" 90-05, which informed Entergy Operations, Inc. of the Palo Verde event, and recomended that utilities with a similar SBCS evaluate their systems for similar potential problems. Although Palo Verde had a foxboro SBCS and Waterford 3 had a Westinghouse 7300 Process Analog Control (PAC) system, the licensee reque;ted assistance from CE on November 28, 1990, to determine if such single failure potentials existed at Waterford 3.
On February 4,1991, CE responded with a list of nine possibilities. Upon reviewing the list and performing a failure analysis on each item, the licensee determined, on February 18, 1991, that component failures in two PAC cards could cause all of the in-service bypass valves to open. The shif t supervisors were infonned of the possibility of such a failure and the inspector noted that the operators'
recomended response to such a failure was ditcussed at the control room shift meeting. Meanwhile, the licensee proceeded to develop a Justification for Continued Operation (JCO) which, by licensee Procedure N0P-015, was to contain descriptions of the problem, the operation of the system, a safety analysis, corrective actions, and reporting requirements.
On February 20, 1991, JC0 91-01 was published, reviewed by the PORC, and approved by the plant manager, as required by procedure. The inspector reviewed the JC0 and noted that the licensee's compensatory action was to
1-insert an additional core operating limit supervisory system (COLSS) penalty of 5 percent to provide additional margin against the occurrence of fuel failure
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due to an inadvertent multiple steam bypass valve opening event. Based on the licensee's safety analysis, the inspector concluded that this action was conservative in view of the fact that Palo Verde ultimately added a COLSS margin of 0.5 percent, and Palo Verde had a steam bypass capacity of 80 percent power (eight valves) compared with 63 percent power (six valves) at Waterford 3.
l The inspector confirmed that the COLSS adjustment was made on February 20, 1991.
The licensee also contracted CE to perform a detailed analysis to determine any further action that would be required.
The licensee estimated that by April 15, 1991, the detailed analysis of the inadvertent opening of multiple steam bypass valves would be completed by CE to detennine what, if any, COLSS penalties would be required in addition to the previously inserted 5 percent penalty to ensure that no fuel failure would result from this event. At the end of this inspection period, the licensee evaluated this condition for reportability under 10 CFR Part 50.73, pending the outcome of the detailed analysis by CE.
It remains unresolved as to whether the plant had been operating outside its design basis (Unresolved Item 382/9103-02).
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t As a minimum, it appeared that 10 CFR Part 50, Appendix A, General Design Criterion 20, may not have been met. The FSAR did not address the inadvertent
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opening of more than one steam bypass valve as a credible anticipated operational occurrence.
3.2.1 Conclusions The licensee's approach to the above SBCS issue appeared appropriate in most respects. The safety significanc2 appeared minimal, given the differences between Palo Verde and Waterford 3.
In view of all the information available to the licensee following the October 20, 1990, Palo Verde event, there appeared to be a long delay from the time the licensee was alerted to the potential condition until February 20, 1991, when compensatory action was finally taken, pending the detailed analysis by CE.
3.3 Blown Packing-from Feedwater Valve FW-163-B At about 5:10 p.m. -(CST) on Sunday, February 10, 1991, the packing blew out of Motor-0perated Valve FW-163-B while the plant was operating at full power.
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Valve FW-163-B is located on the reactor auxiliary building roof adjacent to, and upstream of, Startup Feedwater Regulating Valve No. 2.
The operator determined that approximately 30 gallons per minute of approximately 430*F, hydrazine treated feedwater was being released, most of which was cooled and
, condensed by surrounding structure, and directed to the storm drains.
FW-163-B did not have a backseat and could not be isolated without causing a plant transient. The operators promptly diluted the draining feedwater with normal station tap water to minimize the hydrazine release. The release was subsequently determined to be well below environmental reporting limits.-
Maintenance management responded in a timely manner and procured a leak repair contractor to inject sealant into the packing gland of FW-1638. The process was successfully executed and the release was terminated.
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3.3.1 Conclusions The.0perations and Maintenance Departments' responses to the event appeared timely and effective in minimizing the impact on the plant and the environment.
4.
MONTHLY MAINTENANCE OBSERVATION- (62703)
The station maintenance activities affecting safety-related systems and components listed below were observed and documentation reviewed to ascertain that the activities were conducted in accordance with approved work authorizations (WAs), procedures, Technical Specifications, and appropriate industry codes or standards.
4.1 WA 99000372 On January 15, 1991, the inspector observed the installation and retesting of a new type digital volt meter (DVM) on the plant protection system bistable control panel (BCP) for Channel A.
This werk was to be implemented in accordance with Design Change (DC) 3276. The original DVM had failed in early i
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1989, and no direct replacements could be obtained.. In the interim, a temporary alteration (89-004) was implemented to connect the DVM input to the external DVM jacks so that a portable DVM could be used to accomplish surveillance testing. The work authorized by this WA also removed the temporary alteration.
Even though the new DVM was similar to the oriuinal equipment in appearance, both the wiring and the calibration procedure were
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different, thus a DC was necessary to maintain configuration control.
When the technicians ctarted to install the new DVM connector on the BCP, they noticed a jumper on the old connector that was not addressed in the DC package (DCP). Work was stopped and the design engineer consulted. Although the engineer determined that the jumper would not have affected the operation
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of the new DVM, the procedure should have required its removal. The jumper had been left on the old connector because the procedure was originally written to rewire the old DVM connector. However, the new DVMs were furnished with new connectors, so the technicians took the initiative to use them. Comparison of the new connector with the old connector revealed the extra jumper.
The issue was resolveo by changing the WA to allow use of the new connectors, thus deleting the unwanted jumper.
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After the new connector was installed, the inspector noted disparities between the installed wiring and the DCP. One of the technicians also identified the disparities while verifying his work, as required by the procedure signoff.
The technicians had erroneously assumed that the wires would reterminate on the same numbered pins from which they were removed. The DCP correctly identified different pins on the DVM connector.
The technicians corrected the error.
Thus, the self-checking feature of the procedure had accomplished its intended function.
After the work was completed, a portion of the acceptance test (CI 273064)
required the technicians to perform a 7-step operability test, which involved manipulation of switches on the BCP by an operator while the technicians recorded voltages. When the ic;pector questioned the source document of this retest, (i.e., from what approved procedure the' retest had been taken), the technicians' response indicated that the retest was not part of an approved procedure. Consequently, the design change was retested for operability using a procedure that had not been reviewed by the PORC as required by TS 6.5.1.6.a, and as implemented by Administrative Procedure UNT-007-028, Revision 0, " Design Change Initiation and Review."
Although administrative deficiencies occurred during completion of WA 99000372, the DVM was installed correctly and tested with satisfactory results. Licensee management recognized the problems identified (except the failure to obtain a PORC-approved retest until prompted by the inspector) and took appropriate actions in accordance with their corrective action progrem.
The inadequacy of the DCP to provide for disposition of the unwanted jumper in the DVM connector is a violation of NRC regulations. However, this licensee-identified violation is not being cited because the criteria specified in Section V.G of the NRC's Enforcement Policy wes satisfied.
Failure to obtain a PORC review of the proposed retest was the subject of a Notice of Violation issued on August 1, 1990, with NRC Inspection Report 50-382/90-15.
In response, the licensee
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comitted to implement an improved postmodification and postmaintenance retest:
-program by November 30, 1990, with training completed by December 31, 1990,
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Upon_ reviewing the revised program,- the inspectors noted weaknesses in the
--guidance on PORC review requirements. Failure to review the above retest in accordance with TS 6.5.1.6.a is, therefore, a repeat violation requiring additional corrective actions to ensure that requited PORC reviews are conducted (382/9103-03).
4.2 WA 01069149 On February 6,'1991, the inspector ouserved preparation for work on a leaking
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check valve for Spent Fuel Pool Cooling Pump B.
The inspector reviewed tfe radiation work permit and attended the prejob briefing.
The work package was
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reviewed, including the retest, and was considered adequate for the job.
Posting of the pump room as a contaminated area was completed and several contamination control measures were taken, inclucing covering of floor surfaces with herculit'e and: installation of a containment device to catch leakage from
.the pipe when the flange joint was broken. The tag-out for the work was -
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reviewed and fsund to be' adequate and properly hung..The actual work, which
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included replacing the gaskets for the check valve, was postponed. The inspector was' unable to observe completion of the job. -However, no problems
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were -identified with.the activities inspected.
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4.3 WAs 01069104 and 01072188-
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On February 7,1991, the' inspector observed work on Charging Pump B.
The pump was taken out of service for preventive maintenance, which included changing the ' oil and filter in.the gear box, inspecting and lubricating the motor and pump couplings, and adjusting the crosshead connecting rod bearings. The inspector reviewed the WAs and found that they were properly prepared and
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authorized and were adequate for the work to be perfor.ned. The evolution was
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discussed with the personnel involved. They appeared knowledgeable of the work
'to be performed. The tag-out was reviewed and found'to be adequate. A minor-delay was experienced in obtaining a special spauncr wrench required to check
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the adjustment' on the crosshead connecting rod bearings. This problem was-
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discussed with the cognizant maintenance-engineer. He indicated that ft was not uncommon for tools and other material =to be mispl' aced during decontamination.
The inspector: brought this.to the attention of licensee nanagement as a potential problem. No other problems were identified.
4'.4 Conclusions
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During the implementation of DC-3276 described in paragraph 4.1 above, the inspector noted weaknesses in the developtrent of the DC and the implementing the WA.in terms of' attention to detail, clear identification of electrical connections to be made, and reviews of retests. The licensee recognized several lessons-to be learned-and appropriately entered those let, sons in their corrective action program..As discussed in paragraph 7.4 below, the postmodificationhostmaintenance retest program still'has weaknesses in terms of guidance in assuring that Technical Speicification administrative requirements were met for test procedure reviews und app, ovals.
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MONTHLY SURVEILLANCE OBSERVATIO_N_ (61726)
The inspectors observed the surveillance testing of safety-related systems and components listed below to verify that the activities were being perforn:ed in accordance with the Technical Specifications. The applicable procedures were reviewed for adequacy, test instrwentation was verified to be in calibration,
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and test data was reviewed for accuracy end completeness.
The inspeczors ascertained that any deficiencies identified were properly reviewed and resolved.
Revisio L, " Containment Purge Valve Leak Test"
5.1 Procedure OP-903-112 9 On January 10, 1991, the inspector observed the quarterly local leak rate test of containment purge exhaust Velves CAP-203 and -204 cequf"ed by Technical Specification surveillance requirements ;4.6.1.7.2).
The inspector verified that the test was properly authorized by the shift supervisor and performed by a qualified person using an approved procedure. The procedure nad been changed on January 8,1991, after the penetration for purge inlet Yalves CAP-103 and-104 had been inadvertently overpressurized during the performance of the same test. This event was discussed in paragraph 3.2 obove. The change involved raodification of the pressurization test apparatus and a change in the sequence of valve operations to ensure that tne penetration was brought up to test pressure in a more controlled marr.er, The test was successful'y completed and the velve leakage and updated total containmen-t leakage both met tNe test acceptance criteria.
No prchlems were identified.
5.2 Conclusions
.No violations or deviations were identified during observat on nf the i
surveillance test after Procedure 0F-903-111 was corrected.
However, weaknesses were apparent in the technical reviews of the original procedure change that iciplemented the pressurization apparatus as described 'n paragraph 3.1 of this inspection report.
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OPERATIONAL SAFETv VERIFICATION (717071 The objectives of this inspection were to ensure that this facility was being operated safely and in conformance with regulatory requirenents, to ensure that the licensee's managerrent controls were effectively discharging the licensee's responsibilities for continued safe operation, to assure that selected activities of the licensee's radiological protection programs were implemerted in conformance with plant policies and procedures and in compliance with regulatory requirements, and to inspect the licersee's compliance with the approved physical security plan.
The inspectors conducted control room observations and plant inspection tours and revicwed logs and licensee documentation of equipment problems. Through in-plant observations and attendance of the licensee's plan-of-the-day meetings, the inspectors maintained cognizance over plant status and Technical Specification action statements in effect.
No violations or deviations were identified.
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UFNTIFIED ITEMS (92701, 9302)
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(jLLO90P W MY10V$li _
gl gdj IT1 1289J-01, y
7.1 The insp ct.or conducted a followup inspection on this item in De: ember 1990.
All actions requiring folTowup were completed excyt a review for possible changas in Technical Specifica tion 3.1.3.1.d.
This was docurr)nted in NRC Inspection Report 50-382/90-26. On February 4,1991, the licensee completed their review of Technical Specifics. tion 3.1.3.14 and concluded that the ccrrent Technical Specification was adequate as written, but that the operators
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should monitor the core axial shape index (ASI) more closely when near the O d of core lift and n becomes necessary to rapidly decrease power. When it becomes apparent that a strong negative ASI trend exists and sufficient control mechanisms are not available to stop the trend, the operators should trip the reactor. As the currsnt fuel cycle approaches the end of core life,
~1 the inspector noted that the operators had been made aware of AST trendicg problems as a result of lessons learned from *he trip tha1L oct.arred in 1939 just prior to i.he tTird refueling outage. The licenset W also indicated on intent to continue working w th Combustion Engineering iid licensing to persue possible improvements in thu affected Technical SpecificatRns as part of the Technica. SpecificMion improvement program. This item is closci
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7.2 (Closed) IFl 382/8923-03 On August 3,19Pf, comparent cooling water dry cooling tower Fan IB failed due to a burned motor windint. This i+em was identified as a followup on tne motor vendor's e'*aluation as to cause and poter.ffiwl generic implications.
In a letter dated February 6,1990, Westinghouse E7ettric Corporction responded with a report of the fai7ure nnalysis. The cause :f the winding failure could not be determined. The ver.cor stated that the original design reccids were also reviewed, ant' M problems were identified. The vendor stated that wilen the motor wss minuNctured, it had received a complete engineering test at the factory Lefore it vai shipped to Waterford 3.
Based on the above, and the limited number of winding failures on these motors, Westingnouse concluded that there was no basis for georic concern. The licensee. searched the Nuclear Plant Reliability Data Systeo (NP@S) and found no failures of this type.
Based on the chove, the inspector ccncluded that this was an isolated failure.
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This iten is c~losed.
7.3 Ma,se,q) Violation 3fg2dO1-0,2, J
This viola' don irrvolved an inaceqaate po tmoc'ification retest procedure, Following implementatien of a design cha1ge tha t affected the operability of t
l certain cuntrol roco air ccm'itioning (CRAC) valves. Surveillance Test
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Procedure PE.005-004, Revistun ?, "Contrtl Roorn /,1r Conditioning System l
SLrveillance," was utilized to rete'.t t he valves.
However, the inspector noted l
that Procedare 0 006-004 did not test two of the four valves on which the
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modification was perforr,ed. The liceisee promptly issued and implemented a tete:.t procedum that satisfactorily tested the valves. The 1fcensee I
l tubseciuontly rev'ised Procedure PE-005-004 to test all of the valves during future surveillance testing. This yas done as an enhancerront rather than to l
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meet specific Technical Specification surveillance requireents.
The licensee
also reviewed other surveillance procedures relating to safuty-relato6 ventilation systems for similar conditions and found none. As consnitted to NRC Region IV statf on October 4,1990, the licensee developed and implemented
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improved retest guidance under new Administrative Procedure UNT-005-020 Revision 0, " Post Maintenance Testing." This procedure referenced Nuclear Operations Procedure N0P-014. " Design Changes," which appeared to have an biequate section on postWification testing. The inspector noted that although
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i training was completed as committed, and lead disciplint planners may be aware of Procedure UNT-005-020 requirements, the work package preparation procedure,
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UNT-005-015, Revision 1, " Work Authorization Preparation and Implenentation,"
did not refer to Procedure UNT-005-020 for retest guidelines and requirenents.
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The inspector suggested that the licensee consider adding this important reference with the next change to UNT-005-015.
At the exit interview, the licensee informed the inspector that the change was already in process.
ibis violation is closed.
7.4 _(Open) Violation 382/9015-03
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This violation involved failure of the iicensee to obtain PORC review and approval. of a postmodification test procedure used on the CRAC valves discussed in Violation 382/9015-02 above. The licensee's corrective action was to revise Administrative Procedure UNT-007-02r " Design Change Initiation and Review."
The licensee also coninitted to implewnt an improved postmaintenance retest program, which was completed by issuance of Procedure UNT-005-020 Revision 0,
" Post Maintenance Testing." Although the revision to Procedure UNT-007-028 specifically required acceptance tests to be reviewed by the PORC and approved by the general manager, Proedure UNT-005-020 did not comply with the revision.
Consequently, it appeared that lead discipline planners and system engineers could write a new retest on the postmaintenance test addendum page of a WA and implenent the test without first obtaining PORC-review and general manager
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approval. The violation in paragraph 4.1 of this inspection report could have
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occurred as a result-of this lack of guidance.
The inspector discussed this problem with the licensee. This violation shall remain open until adequate-
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guidance on PORC review of postmaintenance and postmodification testing is l
provided by the licensee.
7.5 (Closed) Unresolved Item 382/9026-03 This item was unresolved until the licensee could determine if the correct model of-General Electric HFA relays were installed in several compartments of l_
Switchgear 3A3-S. The Electrical Mad.nbr.snce Assistant superintendent I
contacted the vendor and verified that, since the relay had a manual reset button, they were the Model -B95F relays as required. He indicated-that the
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label plates on the relays would be corrected as they came due for calibration.
Problem Evaluation /Information Request (PEIR) No. 61613 was submitted to-plant engineering to docunent this detennination and the proposed corrective action. An IFI will be opened to verify that the relays (CS-EREL3A-6G, ACCEREL3A-3G, RFREREL3A-9F, and 4KVEREL3A-11F) have their label plates corrected (382/9103-04). This unresolved itt.n is closed.
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7.6 (Closed) Unresolved item 382/9026-04 This item was unresolved due to an apparent inadequate review of Procedure ME-007-050, Revision 2, " Testing Procedure Electroswitch Control / Latching Relay."
It appeared that the change in the relay testing sequence in Section 8.4 of the procedure was inadvertently altered and did not l
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accomplish what was intended. Discussion with the electrical maintenance assistant superintendent revealed that the in-place (installed in the panel)
cycling of the relay was intentionally deleted from Revision 2 of t!e procedure but agreed that thr.re wa: no difference between cycling the relay in the shop or in the field. A PEIR (No. 61505) was submitted to Plant Engineering because it was questionable whether the cycling of a new or refurbished relay was
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necessary at all since it did not appear to be related to the failure mechanism
being' experienced.- This item.is closed.
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8.
LICENSEE EVENT REPORT (LER) REVIEW (90712)
The following LERs were reviewed in office. The inspectors verified that
reporting requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, and the LER forms were complete. The inspectors confirmed that unreviewed safety
~ questions and violations of Technical Specifications, license conditions, or
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other regulatory requirements had been adequately described. The NRC tracking status is indicated below:
8.1 (Closed) LER 382/90-015. " Inadvertent Control-Room Ventilation Actuation Due to Equipment Malfunction" 8.2- (Closed) LER 382/90-018, " Low Temperature-Overpressure Protection I
Verification Not Performed Per Technical Specifications" 8.3 (Closed)1ER 382/90-020, " Chlorine Gas Release from local Chemical Plant" 9.
ONSITE LICENSEE EVENT REPORT-(LER) FOLLOWUP (92700)
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The following LERs were selected for onsite followup inspection to determine-whether the licensee has taken the corrective actions as stated in the LER and whether responses to the events were adequate and tret regulatory requirements, licensee conditions, and comitments. The NRC trackin: status is indicated i
below:
9.1 (Closed)LER 382/90-006, " Shutdown Cooling System Relief Valve
Setpoint Not 5et In Accordance With Technical SpecifiE h ons Due to
- Procedural Inadecuacy-The inspector verified that the' lift setpoint for Valve SI-406A was properly adjusted during the October 1990 outage in accordance with WA 01059339. The inspector also verified that Mechanical Maintenance Procedure HM-007-006, Revision 3. " Safety injection Relief Valve Test," was issued on October 5, 1990, with the proper setpoints as required by TS 3.4.8.3.
Documentation of
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- i corrective actions was reviewed to ensure that procedure reviews and lessons learned meetings were conducted as conrnitted in the LER. The results were satisf actory. This LER is closed.
10.
PREPARATION FOR REFUELING (60705)
Luring the re;orting period the inspectors reviewed the licensee's preparations
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.for the fourti refueling outage scheduled to start March 15, 1991. A nweting-
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was held on January 24, 1991, with the outage preparation manager und other members of the plant staff to discuss the outage preparation status, major work -
items to be accomplished during the outage, and other items of common interest.;
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The Chief, Project Section A Division of Peactor Projects, Region IV. and the Waterford 3 project manager from the'0ffice of Nuclear Reactor Regulation were also in attendance.
The inspector reviewed selected procedures to be used during the upcoming
- outage, including Refueling Procedures RF-001-001, Revision 1, " Refueling Administration"; RF-001-002, Revision 0, " Foreign Material Exclusion f or-Pefueling Activities"; RF-002-001 Revision 3, Fuel Receipt"; and i
Administrative Procedure UNT-008-030, Revision 8, " Control and Accountability t
of Special.Nuciear Material." No problems were identified.
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On January 9,1991, the inspector observed a portion of-the new fuel receipt inspection. The activities in progress were discussed with the refueling controller. The evolution was well controlled and' performed in accordance with approved procedures, j
11.
EX1T INTERVIEW The-inspection scope and findings were sunrnarized on February 20, 1991, with those persons indicated in paragraph 1 above. The licensee acknowledged the inspectors' findings. The licensee did not identify as proprietary any of the
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material provided to, or reviewed by, the inspectors during this inspection.
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