IR 05000424/1986090
| ML20207B178 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/03/1986 |
| From: | Fillion P, Jape F, Mcelhinney T, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207B159 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM 50-424-86-90, NUDOCS 8611110525 | |
| Download: ML20207B178 (8) | |
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NUCLEAR REGULATORY COMMISSION
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REGION 11
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-Report No.:
50-424/86-90 Licensee:
Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Docket No.:
50-424 License No.:
CPPR-108 Facility Name: Vogtle 1 Inspection Condu ted:
ept ber 22-26, 1986-
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SM'4 Inspectors:
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L. E. Nicholson DitV5igned
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P.J.fillion G/
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Date Signed tf emw tp kcSAbrzscu
///}/76 T. F. McElhinney (/
Date Signed Approved by:
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F. Jape, Sectidfi Chief (/ /
Date Signed Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection involved the areas of preoperational test, witnessing,. test results review, and followup on previously identified items.
Results:
No violations or deviations were identified.
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8611110525 861105 l
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- REPORT DETAILS
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Persons Contacted
' Licensee Employees.
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- R. E. Conway, Senior Vice-President, Vogtle Project Director
- . P. D. Rice, Vice-President, Vogtle Project
- R.-H. Pinson, Vice-President,_ Project Construction
- C. E. ~Belflower, Quality Assurance Site Manager - Operations
- G. Bockhold, Jr., General Manager, Nuclear Operations
- P. R. Bemis, Manager, Engineenng
- R. M. Bellamy, Manager, Test and Outage
- C. W. Hayes, Vogtle Quality Assurance Manager
- A. L. Mosbaugh, Superintendent, Engineering Services W. Chenault, Test Supervisor R. Garrett, Test Supervisor T.:D. Steele, Test Supervisor N. C. Lee, Test Supervisor K. Burr, Lead Test Supervisor J. Aufdenkampe, Lead Test Supervisor J.-Ryan,: Test Supervisor-Other licensee employees contacted included engineers, technicians, operators, mechanics, and office personnel.
Other. Organization H. M. Handfinger, Preop Test Superintendent, Bechtel-M. L. Bagele, Test Supervisor,_Bechtel NRC Resident Inspectors
- J. F. Rogge, Senior Resident Inspector R.lJ.' Schepens, Resident Inspector
- Attended exit interview-2.
Exit Interview The-inspection scope and findings were summarized on September 26, 1986, with those persons indicated in paragraph-1 above. The inspector described the areas inspected and discussed in detail the-inspection findings.
No dissenting comments were received from the licensee.
The following new item was -identified during this inspection.
Inspector Followup Item 86-90-01, Resolve Safety Injection (SI) Block
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When Transferred to Remote Shutdown Panel, paragraph r:
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The licensee did not identify as. proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
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Licensee Action on Previous Enforcement Matters
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This subject was not addressed in the inspection.
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Unresolved Items
Unresolved items were not identified during the inspectio'n.
5.
FollowuponPreviousInspectionItems(92701)
(0 pen)50-424/86-52-01: Review Evaluation of ODR T-1-86-1839.
The NRC inspector investigated the circumstances surrounding a loss of offsite power event that occurred on June 3,1986.
The loss of offsite power was caused by the misoperation of auxiliary protective relays for the reserve auxiliary transformers (RATS).- Due to a wiring error in the Boric Acid System (non-Class 1E panel), 120 volt AC was superimposed on the plant, 125 volt DC non-class 1E distribution system and, metal oxide varisters :
(MOVs) that were connected to the protection circuits in the switchyard were driven into their conducting region. When the MOVs were conducting current, a " sneak circuit" was made up through a low-impedance, high-speed auxiliary relay. -Since the auxiliary relays for both RATS were on the same DC power supply, the misoperation deenergized both RATS and, thus.all offsite power was' lost. Corrective actions proposed by the licensee are:
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Remove the erroneous jumper, i
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Remove any MOVs that are applied in conjunction with low-impedance high-speed auxiliary relays.
There are about 40 such MOVs in the switchyard.
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For all control cables that run between the powerhouse and the switchyard (about214 cables),testtheshieldsforcontinuity, d.
Modify the RAT's protective relay circuit such that a trip signal from the switchyard protection will energize the RAT's lockout relay directly rather than through an auxiliary relay.
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Issue an application guide for MOVs.
Plant personnel stated that a letter to the NRC on this subject will be forthcoming and that the letter will discuss DC power supplies associated with the RAT protection.
This item remains ope e
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6.
IntegratedtafeguardsTestPrdcedureReview(7d304)'
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The NRC inspector reviewed a copy of a prelloinary Or,t procedure for the integrated safeguards and load sequencing tesn'(1-300-01).
Part of this test.is to verify * proper system response to >1oss of offsite power. This preliminary review indicated the procedure was adequate,-;however, the NRC requested a copy"of' the approved procedure for review prior to performing the test.
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No violations or deviations were identified.
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Preoperational: Test Witnessing
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The inspectors observed specific tests being conducted to determine if requirements were being met relative to NRC, requirements such as contained in Regulatory Guide (RG) 1.68 and the Final Safety. Alialysis Report (FSAR).
The following attributes were aidong those virified 'in ti)is, review.
Testswere'psrformedinaccordancewithapprohedObcedures.
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Latest revisions of the approved test procedures were available and in
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use by personnel perfoming the tests.
- Test equipment required by the procedures was calibrated and installed.
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Test data were properly collected and recorded.
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Adequate coordination existed among personnel involved in the test.
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, Test prerequisites were met.
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Proper plant systems were in service.
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Temporary modifications such as jumpers Ure installed and tracked in
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accordance with administrative controls.
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Problems encountered d0 ring testing were properly documented.
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The following tests;were witnessed:
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1-3GJ-01, Essential Chilled 'Jater System (70312)
.The inkpectors witnessed portions of Section 6.5 that performs the operational checkout of the Train "A" Essential Chilled Water dystem. ~ This system is designed to maintain an ambient air temperature below 104*F in the engineered safety features (ESF)
' equipment rooms and the switchgear rooms during operation under accident conditions.
The checkout also verifies system operation from the auxiliary, shutdown panel.
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The inspectors noted a concern that the testing of the essential chilled water system was not being conducted with an adequate
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load.
Appendix A of RG 1.68 states that testing of engineered l
safety features support systems to demonstrate that they meet the
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design requirements should be conducted concurrent with. testing of l
the engineered safety features equipment.
The licensee stated
I that further testing is planned during plant startup by procedure J
1-600-14 Ventilation Capacity Verification.
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1-3S8-01, Reactor Protection System (70305, 70317)
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-The inspectors witnessed portions of Section 6.14 that tested the L
reactor coolant pump trip logic within the ' safeguards test
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cabinet.
The test supervisor in charge was very knowledgeable
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with the test requirements and supporting documents.
Previous-i
. problems encountered during the. performance of this test was discussed with the personnel involved.
This overview of the reactor protection system also included ~ a verification that reactor trip functions actually occurred at the correct setpoints.
c.
Procedure 1-3KJ-05, " Diesel Generator Train "A" Synchronization, Load Rejection, Five Air Starts and 35 Consecutive Starts."
The inspectors witnessed portions of the 35 consecutive starts which are performed in accordance with RG 1.108, Section c.2.a(9).
This test demonstrates the reliability of the Emergency Diesel
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Generator (EDG). The EDG is started and loaded to at least 50% of the continuous rating and operated at this load for at least one hour (35 consecutive times without a. failure) as defined in P.G 1.108 c.2.e.
The EDG was started, manually loaded to 5500 kw and run for an hour.
This test was successfully completed two consecutive times.
On the third start attempt, the EDG failed to-start.
The diesel had not yet obtained the iated speed and
. voltage when the fuel racks went to the closed position.
EDG A had experienced two previous trips of unknown cause on September 17, 1986, during performance of the 35 consecutive start test.
In the two previous trips, the EDG was loaded to 5500 kw and 3800 kvar when the load dropped to zero and the reverse power relay functioned to trip the engine. Georgia Power Company (GPC)
test personnel reviewed information from the third failure provided by a chart recorder and determined that the problem was
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l the hydraulic governor.
GPC decided to bring in the vendor l
representative (Woodward) to assist in resolving this problem.
l The Woodward governor representative reviewed the information and L
concurred with GPC test personnel that the hydraulic governor was not functioning properly. The hydraulic governor was replaced and sent back to the Woodward offices for further testing.to determine the root cause of the failure.
Subsequent conversations with NRC Resident Inspectors indicated that EDG A had successfully completed 11 of the 35 consecutive starts as of September 29, 1986.
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No violations or deviations were noted. \\
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Preoperational Test Results Evaluation'(70329)
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The NRC inspector reviewed the results of Pre-operational Test 1-3PH-01 for Class IE 480 Volt MCCs.. This procedure is essenti' ally for initial power receiving on the Class 1E motor control centers and, it correspond to FSAR Section 14.2.8.1.73.
Results were acceptable.
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No violations or deviationg were noted.
9.
Independent Inspection (92706)
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The inspectors noted from a review of Vogtle's FSdR that the automatic initiation of a safety injection is defeated when control is transferred to the remote shutdown panels.
Paragraph 7.4.3.1.1 of the FSAR states that manual action is required if a safety signal is generated while oerforming a shutdown frota these panels.
This automatic blocking af safety injection is
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currently under review by the NRC with a decision tending., This item will be tracked as Inspector Followup Item (IFI) 86-90-01, Resolve SI block when transferred to remote sh'tdown panel.
u No violations or deviations were identified.
10. Three Mile Island Task Action Plan Followup - Unit 1(4254018)
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This inspection consists of verification that the licensee has
implemented the requirements of NUREG 0737, "Cla'rification of TMI
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Action Plan Requirements" as committed to in the Final Safety Analysis Report (FSAR) or other appropriate documents. Verification consists of one or more of the following attributes, to determine acceptability for
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1the actico item below:
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Program or procedure established
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Personnel training or qualification
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Completion of item v
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Installation of equipment
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Preoperational tests complete
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Equipment calibrated and operable
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Drawings reflect the as-built configuration
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, }e following documents were utilized in performing' tb{si review:
NUREG 0694 TMI Related Requirements for New Operating Licenses
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NUREG 0737 Clarification of TMI Action Plan Requirements FSAR(thru Final Safety Analysis Report
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Amendment 24)
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NUREG 1137 Safety Evaluation Report and Supplements
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~" Auxiliary Feedwater System Automatic Initiation and Flow Indication",
II.E.1.2 (open)
.The :TMI-2 ' action plan requires the ' licensee to p'rovide automatic initiation of_ the auxiliary feedwater system (AFWS) and AFW flowrate indication at the main control and remote shutdown panels...FSAR Sections 7.3.7 -' and 10.4.9 describe that. the AFW system meets the following requirements which are delineated in E.1.2:
Automa*.ic initiation.
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A single failure will not cause the. loss of AFW system function
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Manual initiation can be performed at the main control board and'
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the shutdown or auxiliary feedwater panels
- Testability
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Powered from emergency buses
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Manual capability to initiate the AFW system from the control room-
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and that a single failure will not result in -loss: of the AFW system function The motor driven AFW pumps and valves are _ sequenced on the
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emergency diesel generators Loss of automatic' initiation will not result in the loss of manual
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capability to initiate the AFW system from'the control rcom-Redundant AFW flow instrument channels provided for each steam
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generator Each channel is_ powered from a separate Class 1E power source
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AFW flow' indicators are environmentally qualified
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AFW flow indicators are located at the main control board and at
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the remote panels L
The inspectors conducted a review of the following elementary diagrams to
verify that the power supplies for the automatic iaitiation and flow
indication of the AFW system are' designed as described in the FSAR.
Diagram No.
Title
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1X3D-AA-D028 Motor Control Center 1AAA 1X3D-AA-F17A Motor Control Center 1BBB 1X3D-AA-D03A 4160 V Switchgear l
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IX3D-BC-F08C Discharge Valve 1HV5137 Power Supply 1X3D-AA-E17A BBB16 Switchgear 1X3D-AA-H02A 125 VDC Control Power Supply 1X3D-AA-dO4A
"C" Train Power Supply 1X3D-BC-F04A AFW System 1-1302-P4-003-M01 The-following piping and instrumentation drawings were reviewed:
Drawing No.
Title 1X4DB161-2, Rev. 16 AFW System No. 1302 1X4DB161-3, Rev. 15 AFW System, AFW Pump Turbine Review of the diagrams and drawings by the inspectors indicates that the AFW automatic initiation and flowrate indication are designed in accordance with the applicable requirements and commitments.
The licensee has performed the preoperational testing for the AFW system but the test can not be clo' sed out until some retesting is performed due to modifications to the discharge motor operated valves and flow orifices.
Pending the completion of the preoperational testing and hardware vgrifica-tion, this item will remain open.
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