IR 05000424/1990013
| ML20058L714 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/18/1990 |
| From: | Aiello R, Brockman K, Burnett P, Rogge J, Starkey R, Trocine L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058L711 | List: |
| References | |
| 50-424-90-13, 50-425-90-13, NUDOCS 9008080044 | |
| Download: ML20058L714 (21) | |
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- UNITE 3 STATES
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'o NUCLEAR REGULATORY COMMISslON
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101 MARIETTA STREET.N.W.
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t ATLANTA GEORGIA 30323
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Repor 50-424/90-13 and 50-425/90-13 Licensee: Georgia Power Company
P.O. Box 1295 Birmingham, AL 35201
Docket Nos.:
50-424 and 50-425 License Nos.: NPF-68 and NPF-81
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Facility Name:
Vogtle 1 and 2 Inspection Conducted: May 19 through June 29, 1990 Inspectors: M / @ m t
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J.,Friogge, Senior ResidentMnspector Date Signed mA A
7 /7-/d R d. Aiello, Acting Seniof4esident inspector Date Signed
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R., W 3tarkey, Resident Insppc~ tor Date Signed
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.7-/.7 -9 0 L,-YFocine, Project Engineer Date Signed W
swwW h&9O WBu ett, Regional Inspector Date Signed Approved By: hM eM ~- -- -
7 ~/ 7 * 70 K./E, Brockmsfi, Stction Chief Date Signed
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pivisionofReactorProjects
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i SUMMARY Scope:
This routine inspection entailed resident inspection in the following areas: plant operations, radiological controls, maintenance, surveillance, security, and quality programs and administrative l-controls affecting quality.
Results: Three non-cited violations were identified.
Two of these non-cited violations were in the area of surveillance for failure to perform an adequate TS surveillance resulting in a TS 4.6.1.5 violation (paragraph 3.b.(3)(b)) and failure to perform a power range calori-l.
metric channel calibration resulting in a TS 4.3.1.1 violation
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(paragraph 3 b (3)(e)).
The third violation was in operations for failure to establish an adequate procedure resulting in an i
inadvertent feedwater isolation and subsequent TS 6.7.1 violation (paragraph 3b.(3)(a)).
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ADOCK 03000424 PDC
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No' strengths or weaknesses were identified.
However, the inspector
~ discovered a deficiency card that was. submitted to the US$ 7 days followin!) discovery of the deficiency 3This event was not cited as a potential violation due to the prompt corrective action taken by your technical staff ' and the fact that it was ' an. isolated case.
Deficiency cards '.will-. continue to be closely monitored by the-
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DETAILS 1.
Persons Contacted Licensee Employees J. Aufdenkampe, Manager - Technical Support G. Bockhold, Jr., General Manager - Nuclear Plant G. Frederick, Safety Audit and Engineering Group Supervisor
- T. Greene, Assistant General Manager - Plant Support
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H. Handfinger, Manager - Maintenance
- K. Holmes, Manager - Training and Emergency Preparedness W. Kitchens, Assistant General Manager - Plant Operations R. LeGrand, Manager - Health Physics and Chemistry G. McCarley, Independent Safety Engineering Group Supervisor W. Mundy, Quality Assurance Audit Supervisor
- R. Odom, Nuclear Safety and Compliance Manager J. Swartzwelder, Manager - Operations i
Other licensee employees contacted included technicians, supervisors, engineers, operators, maintenance personnel, quality control inspectors, and office personnel.
- Attended Exit Interview An alphabetical list of acronyms and initialisms is located in the last paragraph of the inspection report.
2.
Operational Safety Verification - (71707)(93702)
The facility began this inspection period with Unit 1 and Unit 2 at 90%
and 100% power respectively.
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Unit 1:
On May 19, 1990, the unit was at 90% power due to a heater drain pump outage.
The unit was restored to 100% power on May 20. On June 22, the unit was manually tripped from 19% power and Mode 3 (Hot Standby) was achieved criticality, entered Mode 1 (Power Operation) y 24, the unit entered due to main turbine vibration problems.
On Ma and tied to the grid.
The unit remained at full power, with the exception of minor power reductions for maintenance, through the end of this inspection period.
Unit 2:
On June 13, 1990, the unit reduced power to 90% due to a heater drain tank level control problem.
The unit subsequently returned to 97% power on June 17, and then initiated a unit coastdown at a boron concentration of 35 ppm in preparation for the September re'ueling outage, 2RI.
On June 28, the unit was manually tripped from 87% power when an MSIV
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developed a hydraulic leak waich caused the valve to close.
The unit remained shutdown to facilitate repairs through the end of this inspection
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a.
Control Room Activities.
Control Room tours 'and observations were performed to verify that
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facility operations were being safely conducted within regulatory requirements.
These inspections consisted of one or more of the
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following attributes, as appropriate, at the time of the inspection.
- Proper Control Room staffing
- Control Room access and operator behavior
- Adherence to approved procedures for activities in progress
- Adherence to Technical Specification Limiting Conditions for Operations
- Observance of instruments and recorder traces of safety related and
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important to safety systems for abnormalities
- Review of annunciators alarmed and action in progress to correct
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same
- Control Board walkdowns
- Safety parameter display and the plant safety monitoring system operability status
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- Discussions and interviews with the On-Shift Operations Supervisor, Shift Supervisor, Reactor Operators, and the Shift Technical Advisor (when stationed) to determine the plant status, plans, and to assess operator knowledge
- Review of the operator logs, unit logs, and shift turnover sheets No violations or deviations were identified, i
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Facility Activities
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Facility tours and observations were performed to assess the effectiveness of the administrative controls establ4hed by direct observation of plant activities, interviews and discussions with licensee personnel, independent verification of safety systems status and LCOs, licensee meetings-and facility records.
During these
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inspections the following objectives were achieved:
(1) Safety System Status (71710) - The inspectors verified that flowpath valve alignment, control and power supply alignments, component conditions, and support systems for the accessible portions of the ESF trains were proper.
The inaccessible portions were confirmed as availability permitted.
(2)
Plant Housekeeping Conditions -
Storage of material and components and cleanliness conditions of various areas throughout the facility were observed to determine whether safety and/or fire hazards existed.
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(3) Fire Protection - Fire protection activities, staffing and equipment were observed to verify that fire brigade staffing was appropriate and that fire alaans, extinguishing equipment, actuating controls, fire fighting equipment, emergency equipment, and fire barriers were operable.
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On June 6, 1990, the inspector observed a routine announced fire drill. The simulated fire occurred in an electrical chase above a battery room on level B of the Unit 2 Control Building.
The drill was well executed.
(4) Radiation Protection - Radiation protection activities, staffing and equipment were observed to verify proper program implementation.
The inspection included review of the plant program effectiveness.
Radiation work permits and. personnel compliance were-reviewed during the daily plant tnurs.
Radiation Control Areas were observed to verify proper identification and implementation.
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Security - Security controls were observed to verify that security barriers were intact, guard forces were on duty, and access to the Protected Area was controlled in accordance with the facility Security Plan.
Personnel were observed to verify proper display of badges and that personnel requiring escort were properly escorted.
Personnel within Vital Areas were observed to ensure proper authorization for the area.
Equipment operability or proper compensatory activities were verified on a periodic basis.
(6) Surveillance (61726)(6.570) - Surveillance tests were observed to verify that approvea procedures were being used, qualified personnel were conducting the tests, tests were adequate to
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verify couipment operability, calibrated equipment was utilized,
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and TS requirements were followed.
The_ inspectors observed portions of the following surveillances and/or reviewed completed data against acceptance criteria:
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Surveillance No.
Title 14235-2 Rev. 2 On Site Power Distribution Operability Verification 14825-1 Rev. 15 Quarterly Inservice SG Sample Valve Test
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14801-1 Rev 6 NSCW Transfer Pump Inservice Test 14802-1 Rev. 5 Quarterly Train "A" NSCW Pump &
Discharge Check Valve IST
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Surveillance No.
Title
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(Continued)
14430-2 Rev. 1 NSCW Cooling Tower Fans Monthly Test 14614-2 Rev. 1 SSPS Slave Relay K608 SI Train "A" Test 14552-2 Rev. 4 Monthly NSCW Flow Path Verification (7) Maintenance Activities (62703) - The inspector observed maintenance activities to verify that correct equipment clearances were in effect, work requests and fire prevention work permits, as required, were issued and being followed, quality control personnel were available for inspection activities as required, retesting and return of systenis to service was prompt and correct, and TS requirements were being followed.
The Maintenance Work Order backlog was reviewed.
Maintenance was observed and/or work packages were reviewed for the following maintenance activities:
MWO No.
Work Description 19002467 Investigate & Repair SGBD Sample Valve Position Indicator 19002711 DG 1B Troubleshooting Plan, Rev. 2 29001981 Install Isolation Valves For DG Jacket Water High Temperature Trip Switches (DCP-90-V2N0166ForDG2A)
On May 21,1990, plant equipment operators were filling the Boric Acid Batch Tank with Reactor Makeu) Water (demin water),
in preparation for batching.
When the cesircd tank' level was reached, the PE0 attempted to close the nakeup valve, A-1228-U4-199, but the valve would not close due to an apparent failure of its diaphragm.
As a result, ti 3 Batch Tank overflowed approximately 11,000 gallons of RMWST water through its overflow line and into the auxiliary building floor drain system. The overflow backed up into the floor drains on 'C'
and
'D' levels of the Unit 1 Auxiliary Building and contaminated approximately 14,000 square feet of floor space.
The resulting HP cleanup and decontamination required several manhours to complete.
Upon investigation, the inspector discovered that valve A-1228-U4-199 had a MWO (A8900374) written on it on March 11, 1989, due to seat leakage.
Through a series of scheduling errors and delays, the valve was never repaired until its ultimate failure on May 21, 1990.
This event was discussed with
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S the Acting Work Planning Superintendent who reviewed the work package history of this valve.
He stated that imnecessary delays had occurred in the processing of this MW0.
He also stated that the schedulers involved had been counseled and, in the future, closer attention would be given by Work Planning Supervisors to the statusing of MW0s.
The inspector had.no further comments.
No vioistions or deviations are identified.
3.
ReviewofLicenseeReports(90712)(90713)(92700)
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In-Office Review of Periodic and Special Reports This inspection consisted of reviewing the below listed reports to determine whether the ;nfermation reported by the licensee was technically adequate and consistent with the inspector's knowledge of the material contained within the report.
Selected material within the report was questioned randomly to verify accuracy and to provide a reasonable assurance that other NRC personnel would have an
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appropriate document for their activities.
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Monthly Operating Report - The Monthly Operating Reports and their revi.ons, dated May 11 and June 7,1990 were reviewed.
The inspector had no coments.
The VEGP 1989 Annual Report - Part 2 was reviewed. The inspector had no comments.
i Special Reports:
(1)
" Loose Part Detection System Inoperability."
The Reactor Verel Upper Channel of the Loose Part Detection System was inoper ble when Unit 1 entered Mode 2 on April 15,
1990.
Troubleshooting revealed that the control panel was operating properly and that the most likely cause of the i
malfunction was in either the sensors attached to the reactor head or a pre-amplifier located about 6 feet away.
These non-safety related components were unavailable for inspection while the reactor was at power but will be inspected after the unit makes its next Mode 5 entry. The inspector has no further
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Comments.
(2)
"VEGP Unit 2 Third Supplement To Summary Repert of Startup Test Program."
This third supplement to the initial startup report addresseo the remaining startup test that was incomplete at the tii.T of submittal of the initial startup report.
The only remaining
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testing to be completed consists of cold post-test readings which require the plant to be in a cold shutdown condition. The
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l readings will be performed during the Unit 2 refueling outage under Procedure 2-600-11.
The irspector has no1further comments, b.
Deficiency Cards and Licensee Event _ Reports
= Deficiency Cards and Licensee Event Reports were reviewed for potential generic impact, to detect trends,.and to determine whether corrective actions appeared appropriate.
Eunts which were reported pursuet to 10 CFR 50.72 were. reviewed. following occurrence, to detern m if the Technical Specifications and other regulatory requit&.:',s were satisfied.
In-office review of LERs-may result in further followup to verify that the stated corrective actions have been completed. or to identify violations in addition to those described in the LER.
Each LER was reviewed for enforcement action-in accordance with 10 CFR Part 2. Appendix C, and where the violation was not cited-the criteria specified in Section V.A or V.G.1 of. the Enforcement Policy were satisfied.
Review of DCs was performed to maintain a realtime status of deficiencies, determine regulatory compliance, follow the licensee corrective actions, and assist in the closure of the LER when reviewed.
Due to the numerous DCs processed only - those DCs which resulted in enforcement ection or further inspector followup with the licensee at the end of the inspection are listed below.
The DCs and LERs denoted with an asterisk indicates that hactive inspection occurred following the event and prior to receipt of the written report.
f (1) The following Deficiency Cards were reviewed:
(a) DC 1-90-246, " Licensed Operator Confirmed Positive On Fitness For Duty Test."
On May 11, 1990, the plant General Manager was-notified of a confirmed positive ' test for THC involving a licensed reactor operator.
The test was a result of the licensee's random Fitness For Duty screening program.
The badge'of the involved licensed operator was promptly deactivated and the individual was escorted out of the protected area. The inspectors will continue to monitor the effectiveness of the Fitness For Duty program.
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(b) DC 1-90-256, "DG 1B Automatically Trips Due To Incorrectly Calibrated Jacket Water Temperature Switches."
On May 23, 1990, I&C installed three recently calibrated CALCON jacket water temperature switches on the IB DG.
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These switches were the first to be installed after being calibrated using procedure 22981-C, Calcon Pneumatic Temperature Sensor Calibration, Rev. O.
After a normal start, the DG tripped af ter approximately 60-90 seconds.
Subsequent DG testing indicated that the newly installed o
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jacket water temperature switches -were the cause of the
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The switches were ' replaced with the original-
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switches which had been removed earlier and the DG tested t
satisfactorily.
The suspect switches were sent to an
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independent laboratory for evaluation.
This event will be i
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further followed up when submitted as a Special Report.
(c) DC 1-90-260, " Response Time For CREFS Actuation Did Not
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include DG Sequencer Loading Delays Nor The Time For Fans To Pressurize Control Room."
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On May 29, 1990, while perfonning > ESFAS Response Time Summation, procedure 54800-1 and 54800-2, the licensee discovered that for CREFS ' actuation, DG sequencer
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determining total. actuation response time. Also, the time required for.the fans to pressurize the control room to y
1/8"' positive pressure had not been determined.
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c circuitry to be inoperable and entered the action statement
for TS 3.3.2, Table 3.3-2. Action 27. A Regional Waiver of Compliance was granted on June 6,.1990, to allow the licensee to stop all. CREFS fans'. to perform the testing
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necessary to collect the _ required data.
Testing 'was
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completed on June 7,1990,_ thus completing the -utilization
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of the compliance waiver.
This event will be further
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'-90-264, " Security System Anomaly Could Allow
..wcected Access.To3A Vital Area."
The ser.urity system vendor advised the licensee by a ler" dated November 1, 1989, of an anomaly that exists ir
' ant security system that could -possibly allow
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Compensatory measures t
implemented until June ;1,1990, when the full
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.f the vendor's information was realized.
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.ne st. urity system.
This_ event will be further
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tollowed up by NRC security inspectors.
X (e) DC 1-90-266 " Failure To Perform ILRT On Airlocks."
h June 1, 1990, during review of the ILRT report for the
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second refueling outage, the licensee discovered that the six months test interval for containment airlocks per TS 4.6.1.3.6.1 was exceeded.
Two instances for the emergency airlock and one for the personnel airlock were reported.
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This will receive further followup when submitted as a LER.
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(f) DC 1-90-270, " Sodium Hydroxide Intrusion Into The RCS."
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On June 1, 1990, it was discovered that sodium hydroxide intrusion had occurred during a-containment spray system
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surveillance.
This deficiency did not result in a ;
violation-of TS en this occasion; however, the potential
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existed due to die _ ability of Na-24 to mask -iodine
.t a th ity.
The inspectcr discovered that this DC was not t
NhsitSd in a timely manner as defined in administration-L Recedure 150-C, This event was not cited due to the prompt corrective action, via a memorandum to all department menwen regarding DC processing, taken by the
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Manager - Technicai Lopport and due tc the fact that it was an isolated case.
Deficiency Cards will continue to be m
closely monitored by the resident inspector.
(g) DC 2-90-064, " Manual Reactor Trip Caused By MSIV 0-Ring Failure on Hydraulic Line" On June 29, 1990, operators manually tripped the Unit 2-
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reactor when they discovered a hydraulic fluid leak in a l
loop #3 MSIV which caused the valve to close.
This event i
will be further followed up when submitted as a LER.
(2) The following LER was reviewed and is ready for ' closure
>v pending verification that the licensee's stated corrective actions have been completed.
50-425/90-03, Rev. O, " Trip Of Heater Drain Pump Results In c
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_ Exceeding The Reactor Power License Limit."
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On April 1,1990, a power excursion resulted in' Unit 2 exceeding the maximum power ' level'(3411 megawatts thermal) specified in 3: (
Operating License NPF-81, Section 2.C.(1).
The power excursion-M occurred when a condensate pump was started after a heater drain pump tripped on HDT low-low level. The pump start caused cooler feedwater flow to the steam generators.
The' corresponding cooldown of the primary system resulted in reactor pawtr peaking-at 105.2% of rated thermal power (3590 megawatts thermal).
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subsequent review of data-indicated that reactor power remained 4~ *
above 100% for approximately nine minutes. -An overpower rod block and an automatic turbine runback occurred as a ' result-of the power excursion.
Operators responded by manually inserting P
control rods, borating, and decreasing turbine load until i
reactor power was stabilized at 90%.
The low-low level in the
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s HDT occurred following a HDT high level and subsequent isolation l'
of-inputs to the HDT.
The-high level occurred as a result of
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the level dump valve failing to automatically open.
The valve failed to open because a manual actuation pin for the valve had
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been inappropriately placed into the manual position.
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inspector questioned the adequacy of the corrective action e
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working on or around equipment was not addressed. The licensee-p_lans to revise this LER.
'(3). The following LERs were reviewed and closed.
(a) 50-424/90-09, Rev. 0, " Inadvertent Feedwater Isolation Due
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To Procedural Inadequacy."
On April 13, 1990, with the unit in Mode 4, an unplanned
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Feedwater Isolation occurred during the performance' of-
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procedure 14703-1,
" Reactor Trip Bypass Breaker Undervoltage Trip".
The FWI resulted in closure of the bypass.feedwater' regulating valves which were open to.
-support condensate long-cycle recirculation.
The steam generators were being supplied by auxiliary feedwater.-
Operator action was taken to reopen the BFRVs and r
reestablish long-cycle recirculation.
The cause of the event was procedural inadequacy.
Technical
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Specification 6.7.1 states that written procedures shall be-
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established, implemented, and maintained as deliniated'in'
Appendix A of Regulatory Guide 1.33, Revision 2. February
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1978.
Restoration steps in procedure 14703-1 1.ncorrectly
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directed that the' Mode St. lector switch, located on the-Solid State Protection System panel, be placed in OPERATE and the Input Error Inhibit switch be placed in NORMAL. An investigation determined that this sequence was incorrect l
and resulted in the FWI.
Corrective actions included.
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revising procedure 14703-1 to require that the Input. Error Inhibit switch be placed in NORMAL, and _the FWI signal be-reset, prior to placing -the Mode Selector' switch in
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OPERATE.
This licensee-identified violation is not being cited because criteria specified in Section V.G.1.if the.
NRC Enforcement Policy were satisfied.
In-order to track'
this item, the following is established.
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NCV 50-424/90-13-01, " Inadvertent FWI Due To A' Procedural
Inadequacy Resulting In A TS 6.7.1 Violation - LER l
1-90-09."
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(b) 50-424/90-10, Rev. O, "Im) roper Recording Of Data Leads ~To
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Inadequate Technical Spec'fication Surveillance."
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On April 18, 1990, during a review of completed procedure L
14000-1, " Operations Shift And Daily Surveillance Logs",
the oncoming Unit Shift Supervisor found that on April 17, l
1990 and April 18, 1990, containment level C temperature L
had been recorded from ERF computer point T7502 which had p
been reading erroneously low since April 11, 1990.
This-L low reading caused the overall containment average L
temperature, comprised of the average of the Level 2, l
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Leve'l B, and Level C' temperature readings, as per Technical
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Specification 4.6.1.5, to te? 1rroneously-low.
A review of
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surveillances performed on v il-12 - 16, 1990, determined that an alternate indication had been utilized and these
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results were satisfactory.
The' surveillance on April 11',
1990, was performed prior to the failure of computer point
T7502.
The cause of this event was a cognitive personnel
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error by the on-duty Reactor Operator and Unit. Shif t -
Supervisor in that they recorded and reviewed data from a malfunctioning indication.
In addition, the USS on duty on-April 12, 1990, failed to ensure adequate turnover of
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information to prevent use of the malfunctioning, computer point.
The USS and R0 responsible for -the inadequate
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surveillances were counseled regarding the importance of
attention to detail in the performance and review of surveillance procedures.
The USS on duty on April 12, l
1990, was counseled regarding-the need to ensure adequate j
turnover of information to the oncoming shif t.
This.
licensee-identified violation is not being cited because'
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criteria specified in Section V.G 1 of the NRC Enforcement
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Policy were satisfied.
In order to track this item,.the J
following is established.
NCV 50-424/90-13-02, " Failure To Perform An Adequate TS Surveillance Resulting In A TS 4.6.1.5 Violation - LER
.l 1-90-10.
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(c)*50-424/90-11, Rev. O, "A Manual Reactor Trip Due To
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Inadvertent Closure Of Main Feed Regulating Valve."
On April 25, 1990, the Unit I reactor was manually tripped
L due to decreasing level in SG No. 2.
Prior to the trip, a l
" Steam Generator 2 Flow Mismatch Alarm" annunciator was-
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-received and feedwater flow to SG No. 2 was observed to be
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L decreasing rapidly.
The B0P Operator attempted to increase l
the feedwater flow by increasing the demand signals to both
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L the Main and Bypass Feedwater Regulating Valves to No. 2
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Huwever, SG. level continued to fall which forced l'
initiation of the reactor trip.
The Auxiliary Feedwater L
System actuated as designed to maintain SG levels and the unit was stabilized in Mode 3.
Subsequent investigation
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indicated that MFRV No. 2 had closed and caused the event.
The MFRV apparently closed when workers installing i ;;
insulation on the MFRV inadvertently bumped into and
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mispositioned the local control levers located on the side
of the valve positioner.
Hispositioning of the local
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control levers interrupted the control air supply to the valve positioner.
Since the local control levers are not used for normal operations, they were removed to prevent recurrence of this event.
The inspector has no further comments.
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(d)
50-425/89-13, Rev.
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" Flood Barrier Removal Leads To Auxiliary Feedwater Inoperability."
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_ Technical Specification 3.7.1.2 requires that three
_ independent steam generator AFW pumps and associated flow paths be_ operable _in Modes 1, 2 and 3.
On March 30,_1989,
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plant personnel were conducting a-routine walkdown when they found a flood protection barrier removed from the wall-
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between the AFW discharge piping room and the Turbine Driven AFW pump room. 'This resulted in the AFW pump being.
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declared ' inoperable and the action statement of TS 3.7.1.2 being' entered.
The barrier was expeditiously replaced and the TS action statement was exited.
The cause of this event was personnel error in that the barrier was removed
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without ' the ' proper review and approval.
Work had been
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performed on a check valve in the discharge piping room.
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.When a functional test was performed on March 23, 1989, the
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existence of a flood barrier and precautions to be observed were not addressed by those requesting the test or by those implementing the work order.
As corrective action, a sign was installed near the flood barrier and information was added to the equipment file advising of the flood barrier's existence.
Also, plant operators were advised of this event by placing a copy of this LER in the Operations Required Reading Book.
The i_nspector has no further comments.
(e)
50-425/90-04, Rev. O
" Personnel Error Leads To Missed
Calorimetric Channel Calibration."
On April 111, 1990,- the-Unit Shift Supervisor for the oncoming : night shift wu reviewing the Unit Shift Supervisor's Log.
At 5:15-p.m. - EST, he noticed-that: the Power Range Calorimetric Channel Calibration had not been
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performed since 11:45 a.m.
EST, on April' 10, 1990.
Technical Specification 4.3.1.1, Table 4.3-1 I tem _2a,
requi es that this calibration be performed daily for the-power range neutron flux high setpoint.
The USS on duty was advised of this and personnel began to perform the necessary calibration. The surveillance interval (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I
plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. grace period) expired at 5:45 p.m. EST, and
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the unit entered operation under TS 3.0.3.
At 6:36 p.m.
EST,- the Calorimetric calibration was satisfactorily completed with no adjustments ' required.
The unit
. subsequently exited TS 3.0.3.
The cause of this event was a cognitive. personnel error by the USS on duty, resulting in a failure to comply with Procedure 14000-C, " Operations Shift And Daily Surveillance Logs." The USS was counseled regarding the importance of compliance with TS require-ments.
This licensee-identified violation is not being cited because criteria specified in Section V.G 1 of
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the NRC Enforcement Policy were satisfied.
In order to
track this item, the following is established.
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NCV 50-425/90-13-01, " Failure To Perform A Power Range
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Calorimetric. Channel. Calibration Resulting In A TS 4.3.1.1
.
Violation - LER'2-90-04."
(f) 50-425/90-05, Rev. 0, " Computer Point Failure Results In
Exceeding The Reactor Power License limit."
Q-On April 29, 1990,. Proteous computer point F0-424A (SG #2
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feedwater flow) was discovered to be reading lower than control-board indications.
Since this computer point provides input to the computer calculated calorimetric
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power indication, F0-424A was promptly removed from the-Proteous scan.
This resulted in an increase of indicated reactor power to 3411 megawatts thermal. While removal of
computer point F0-424A only brought indicated reactor power.
Up to 100% rated thermal power, a subsequent review of
.
computer data indicated that actual reactor power had
slightly exceeded 100% beginning on April 27, 1990.
It is
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estimated that reactor power averaged 100.5% RTP for both a 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and a 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> interval until discovery of the-computer point failure.
A computer input card.for F0-424A was replaced and the computer point was verified to be indicating correctly.
Corrective action-to prevent recurrence included tightening the cluster limit for
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acceptance of the feedwater-flow input. values as good data.
The inspector has no further comments, g
b (g) *50-425/90-06, Rev. O, " Personnel ~ Error Leads To Unit lj Operation Per Technical Specification 3.0.3."
l On May 1, 1990, a Chemistry foreman advised the Unit Shift
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Supervisor that work was required on the -Containment
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Atmosphere Gaseous Monitor 2RE-2562C.
The USS authorized the work but specified that 2RE-2562C and the Containment Atmosphere Particulate Monitor 2RE-2562A, could not both be
taken out of service simultaneously.. The foreman instructed a technician to proceed with the work and at 2:03 p.m. EST, the Data Processing Module was taken to
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purge, which effectively rendered both 2RE-2562C and 2RE-2562A inoperable.
The USS was alerted to 'this
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condition when a spurious high radiation alarm prompted him to investigate the work in progress. Since the TS action statement does not clearly address these two monitors being out of service simultaneously with another leakage detection system which was already out of service, the USS entered TS 3.0.3, which requires unit shutdown to commence within one hour.
The cause of this event was a cognitive personnel error by the Chemistry foreman.
He failed to
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a recognize that the work being performed would render both channels inoperable and he failed to' notify the USS that
o-both channels would be inoperable. - The foreman has been l
counseled. The inspector has no further coninents.
(h)*50-425/90-07. Rev. O, " Relay Failure Results In Reactor Trip."
On May 6,1990, control room operators f received trouble-alarms indicating closure of-Main Steam Isolation. Valve 2HV-3026A and a resultant low-low water level in'SG #3. An automatic reactor trip ensued.
The. direct cause of this event was the closure of the MSIV, which resulted in.the reactor trip when SG #3 reached its low-low water level
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-setpoint.- An investigation of the MSIV controls found that the air solenoid to the' valve's hydraulic pump was not energized so that air pressure was not available to drive i
the pump motor.
Further investigation revealed: that the i
AX1 relay had failed. The AX1 relay energizes both the air j
supply solenoid and the hydraulic dump solenoid to hold the
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MSIV open, When the AX1 relay failed, the loss of hydraulic-fluid pressure to the valve. caused the MSIV to close.
The failure of the relay was the root cause for l
i this event. The failed relay was replaced and the MSIV was
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tested and verified to operate satisfactorily.
The j
inspector has no further questions.
j Three non-cited violations were identified.
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4.
.Incore-Excore Nuclear Instrument Correlation (61705)
j The ' licensee has changed the method of performing the incore-excore nuclear instrumentation correlation surveillance ' required by TS 4.3.1.1'
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. (Table.4.3-1, Item 2.a. Note 6),
(The method is not described in the TS
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or the FSAR.)
The new method was developed by a consulting firm, which j
' has adapted it to Vogtle by analysis of data from both units.
The new method will allow the correlation to be updated whenever a full core flux l
map is perfonned and eliminates the need to deliberately perturb the axial j
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power distribution in order to perform multiple flux maps at: many j
i, different incore axial offsets.
,
i On June 7,- 1990, a region-based inspector attended the consultant's l
presentation of the correlation method and its justification to the plant
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reactor engineering staff.
The presentation was followed by a Plant
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Review Board meeting, at which the procedure chnge to implement the new l
method was reviewed and approved.
A copy of the final, proprietary, consultant's report will be mailed-to the NRC Region 11 office by the licensee.
Following discussions with licensee engineers and the consultant, the inspector had no further questions.
The results of this new surveillance
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method' will. be reviewed as part of the regional-initiative inspection program.
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No violations or. deviations were identified.
5.
Actions on Previous' Inspection Findings - (92701)(92702)-
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a.
(Closed) Violation 50-424/89-13-01, " Failure to Take Adequate
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Corrective Actions For Previous Inspection Findings."
This violation identified -three examples of inadequate corrective
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t action which violated the requirements of 10 CFR 50, Appendix B, criterion XVI.
The three examples were: (1) Failure of operations
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personnel to read the required reading book in a timely manner; (2)
Failure to cover the ute of the TS cooldown limit curve in i
requalification: training for licensed operators; and, (3) Failure to
- i properly-maintain FSAR copy 125.
Each of thete examples have' been addressed by-the licensee and adequate corrective actions have been
taken. The inspector has no further comments.
r b.
(Closed)-Violation 50-424/89-13-02, " Failure To Follow Procedure."
r Three examples were cited for failures to follow procedures which violated the requirements of 10 CFR 50, Appendix B, Criterion V.
In
the first example, Shift Supervisors had failed to document the
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monthly review of the Information Tag Log.- On April 25,_1989, Procedure 10009-C, Operator. Aids, was revised to make the Unit Operations Superintendent responsible for ' the performance of-the
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Information Tag review, rather than the Shift Supervisor on shift.
.The responsibi'ity for the performance of the review was removed from the operating' crews and was-placed on non-shift personnel.
The
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second_ example stated that the licensee had failed to
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perform / document the weekly review of the Annunciator Status Control
- Log for the weeks ~ of January 20 and 27,1989.
The individuals
-responsible were counseled and an Interoffice Correspondence letter from Operations management was placed in the Operations Reading Book
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to make licensed operators and su)ervisors aware of the violation and
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the importance of procedure compliance.
In the third example, the-licensee issued' changes to As-Built Notices 'on white paper rather
'than yellow paper as required.
For corrective action, procedure 00101-C, Drawing Control was revised to allow for distribution of.
J advanced copies of As-Built Notices on white paper prior to filming
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and normal distribution.
The inspector was satisfied that adequate l
. corrective : actions were taken on all three examples and has no further comments.
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(Closed)
IFI 50-425/88-79-02,
" Verify Operations Department
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Comitment To Review Unit 2 Procedures Against Unit 1 Procedures For Omissions."
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u As a result of the identification-of -unjustifie'd discrepancies
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between Unit 1.and Unit 2 nuclear operating. procedures, the licensee.
- conrnitted to correct the noted omissions and = to perform additional procedural evaluations to detect similar problems, in order-to
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verify-corrections to these previously identified discrepancies,-the
inspector reviewed the following operating procedures:
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13001-1, Rev.12, Reactor Coolant System Filling and Venting,
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dated October-10, 1989.
13001-2, Rev. 4 Reactor Coolant System Filling and Venting,
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dated October 10, 1989.
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14804-1, Rev. 7, Safety Injection Pump Inservice Test, dated
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February 25, 1989.
14804-2, Rev. 2, Safety Injection Pump Inservice Test, dated
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March 14, 1989.
14825-1, Rev.13, Quarterly Inservice Valve Test, dated February
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9, 1990.
14825-2, Rev. 4, Quarterly Inservice Valve Test, dated February,
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9, 1990.
j 14980-1, Rev.18, Diesel Generator Operability Test, dated
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February 5, 1990.
l 14980-2, Rev. 4, Diesel Generator Operability. Test, dated
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February 5, 1990.
The discrepancies. listed in NRC Inspection Report Nos. 50-424/88-61 and 50-425/88-79 were corrected; however, the following unjustified; discrepancies were noted:
13001-2, requires Checklist 4, Paragrap(h 4, Item 4 -of procedure
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valve 2-1201-U4 102 pressurrizer safety valve PSV-8010A loop seal drain) to be closed and the same item on the Unit 1 procedure (13001-1) requires valve.1-1201-04-102 (pressurizer safety valve PSV-8010A loop seal drain) to' be' locked closed.
P&lD 1X4DB112, Rev. 22, Reactor Coolant System, indicates that
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this valve should 'be normally closed and does not indicate locked closed.
Operations reviewed the ' requirements for locked.
valves and datermined that a revision was required to procedure 13001-1. Procedure 13001-1, Rev.13 was approved on April 3, 1990, and changed 1-1201-U4-102 from " locked closed" to
"cl os ed".
This change agreed with the P&ID 1X4DB112 in not requiring valves to be locked that have a down stream second isolation valve.
No changes were required to the drawings or to procedure 13001-2.
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- 16 f-Additional licensee corrective actions included: (1)A_reviewofall E0Ps, A0Ps, U0Ps, SOPS, and lineups; (2) Correction of any identified j
problems; and, (3). The establishment of a policy to revise both-
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Unit 1 and Unit.2 procedures when one is being reviewed for revision to ensure that conformity between units is maintained.
The inspector has no further comments.
d.
(Closed) IF1'50-425/89-26-02, " Review Licensee Response To Why The ARVs Were Challenged. Why Auxiliary Feed Pump Bearing Temperatures Alarmed, And Improvements Of The Post Trip Review Process For.,The Reactor Trip On July 26,.1989."
The Proteous high alert and high alarm setpoints for the motor driven AFW pump motor bearing temperatures have been raised to 170 degrees F
, per Computer Software Change Requests T
and 185 degrees F, respectively(Unit 2).
89-0053 (Unit 1) and 89-0054 These CSCRs were installed
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and functionally tested on December 20,1989, and December 22, 1989,
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respectively.
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After detailed review of Reactor Trip Report 2-89-008, Event Report l
2-89-14, and applicable procedures, the licensee has concluded that J
the ARVs. for SGs #1 and #4 operated prematurely and the probable causes were improper setting of the ARVs potentiometer.and/or f
calibration. drift of the associated pressure transmitters.
Procedure L
12004-C. Mode l' (Power Operation), was revised to provide' a procedural step to place the ARVs in proper standby alignment for
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power operation.
The reactor trip review procedure 10006-C was revised to improve the review analysis of trip events by establishing a process whereby data = collection and review is performed prior to -
declaring a cause of the event.
.The inspector has no further coments.
i 6. -
Release From CAL
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Confirmation of Action Letter, CAL-50-424/90-01, dated March. 23,1990,
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specified certain matters that were agreed to be completed by GPC as a result of the March 20, 1990, Site Area Emergency event.
In a meeting conducted on April 9,1990, GPC management briefed NRC management on their event critique results and the resultant short and long-term corrective actions.
All items to which the licensee comitted were specified in a
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GPC letter dated April 9,1990.
An NRC letter dated April 12, 1990,
which referenced 'both the April 9,1990, meeting and letter, confirmed the
satisfactory -resolution of CAL item number 1 and documented the Regional Administrator's concurrence that appropriate corrective actions had been
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taken and that Vogtle Unit 1 could return to power operation.
This
release from CAL item number 1 was also documented in NRC Inspection Report Nos. 50-424/90-05 and 50-425/90-05.
One of the commitments referenced in the April 9,1990, letter was the licensee's agreement to finish reviewing the event review team's I
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i long-term recommendations ~ and to transmit a summary and schedule for the completion of corrective actions to the NRC no later than May 15, 1990.-
By letter dated May 14, 1990, the licensee submitted a summary of the corrective actions resulting from the Site Area Emergency.
In -this
. letter, the licensee committed to revise the maintenance procedures for CALCON temperature switches to include the lessons learned from laboratory
- testing - by May '15,.1990; to clean and calibrate all jacket water high temperature switches.using these revised procedures by May 31, 1990; and to-clean and calibrate other non-essential-trip temperature switches by.
the end of the 'next refueling outage for the associated unit.
The documented the licensee's actions licensee's' letter dated June 22, 1990, '
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regarding this commitment.
Although the licensee was released from CAL item number 1, CAL item numbers 2-5 remained applicable and were not relieved by the NRC letter dated April 12, 1990. These items included the following issues:
quarantining of equipment involved in the incident by the IIT;
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preservation of records or damaged equipment that may be related to
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the March 20, 1990,- event and any surrounding circumstances that could assist in understa_nding the event;.
ensuring the availability of personnel with knowledge of the event
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or its causes for interviews by the IIT; and preclusion of interference with the IIT by._ licensee or third party
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investigations.
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The licensee was fully responsive with regard to quarantined equipment, l
preservation of records or damaged equipment that may have been related to-the event, availability of individuals for questioning,- and conduct of
separate investigations.
The licensee also has a program in place to
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ensure that all records related to the March 20, 1990, event shall be retained for at least two years whether or not required to be retained by e
regulation or license condition.
Based upon the information provided by the licensee and upon the licensee's actions regarding the remaining CAL
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issues, the licensee was. released from CAL-50-424/90-01 on July 20, 1990.
L 7.
Exit Interviews - (30703)
The inspection scope and findings were summarized on June 29, 1990, with l
those persons indicated in paragraph 1-above.
The inspectors described E
the areas inspected and discussed in detail the inspection results.
No l
dissenting conunents were received from the licensee.
The licensee L
identified as proprietary the consultant's report regarding the NI i,
correlation surveillance.
Region based NRC exit interviews were attended during the inspection period by a resident. inspector.
This inspection l
closed two Violations, two Inspector Followup Items, and eight Licensee l
Event Reports. The items identified during this inspection were:
l-l; NCV 50-424/90-13-01, " Inadvertent FWI Due To A Procedural Inadequacy Resulting In A TS 6.7.1 Violation" - paragraph 3.b.(3)(a).
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NCV 50-424/90-13-02, " Failure To Perform An Adequate TS Surveillance m
Resulting In A TS 4.6.1.5 Violation" - paragraph 3.b.(3)(b).
g NCV 50-425/90-13-01, " Failure To Perform A Power Range Calorimetric Channel Calibration Resulting In A TS 4.3.1.1 Violation" paragraph 3.b.(3)(e).
['
r 8.
Acronyms And Initialisms
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AFW Auxiliary Feedwater System
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A0P Abnormal Operating Procedure
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ARV.
Atmospheric Relief Valve BFRV
. Bypass Feed Regulating Valve B0P Balance of Plant su
CAL Confirmation of Action Letter
CDT Central Daylight Time
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CFR Code of Federal Regulations CREFS Control Room Emergency Filtration System t
CSCR Computer Software Change Request CVCS Chemical & Volume Control System DC Deficiency. Cards l
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DCP Design Change Package DG Diesel Generator DPM Disintegrations Per Minute
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E0P Emergency Operating Procedure
ERF Emergency Response Facility ESF Engineered Safety Features
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ESFAS Engineering Safety Features Actuation System-EST Eastern Standard Time F
Fahrenheit FSAR Final Safety. Analysis Report FWI Feedwater Isolation
GPC Georgia Power Company HDT Heater Drain Tank
HP.
Health Physics HV High Voltage HX-Heat Exchanger IFI Inspector-Followup Item IIT Incident Investigation Team ILRT Integrated Leak Rate Test
.IST Inservice Testing LC0 Limiting Conditions for Operations LER'
Licensee Event Reports
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MFRV Main Feed Regulating Valve MSIV.
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MWO Maintenance Work Order Na Sodium NCV Non-cited Violation NI Nuclear Instrument
NPF Nuclear Power Facility a
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NRC Nuclear Regulatory Commission NSCW Nuclear Service Cooling Water System PEO.
Plant. Equipment Operator
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PSV l Pressurizer Safety Valve QC-Quality Control RCS-Reactor Coolant System Rev:
Revision RMWST-Reactor Makeup Water Storage Tank l
R0 Reactor Operator-
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SGBD-Steam Generator Blowdown SI Safety Injection System S0P Standard Operating Procedure
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SSPS'
Solid State. Protection System
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.THC Tetrahydrocannabinol TS Technical Specification UDP-Unit Operating Procedure USS Unit-Shift Supervisor
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VEGPL Vogtle: Electric Generating Plant 2R1 Unit 2 First Refueling Outage
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