ML20044G223
| ML20044G223 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/14/1993 |
| From: | Balmain P, Brian Bonser, Skinner P, Starkey R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044G212 | List: |
| References | |
| 50-424-93-09, 50-424-93-9, 50-425-93-09, 50-425-93-9, NUDOCS 9306020196 | |
| Download: ML20044G223 (6) | |
See also: IR 05000424/1993009
Text
{{#Wiki_filter:. - [/* %g UNITED STATES f NUCLEAR REGULATORY COMMISslON 2 ~ (#gn REGION il l 101 MARIETTA STREET, N.W.
%{{I C ATLANTA, G EORGI A 30323 ,uj ..... Report Nos.: 50-424/93-09 and 50-425/93-09 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 Facility Name: Vogtle 1 and 2 Inspection Conducted: April 18, 1993 - May 8, 1993 Inspector: h6 m / f./.3 9'.F p B. R. Bonse M enior Resident Inspector Date Signed 0 n/ 5./3 93 R. D. Sta , Resident Inspector Date Signed h 6 /3-92 o P. A. Ba) apsident Inspector Date Signed Approved by: BN / d~ I /'/ 73 P. Skinner, Chief Date Signed Reactor Projects Section 3B Division of Reactor Projects SUMMARY Scope: This special inspection conducted by the resident inspectors concerns a safety injection which occurred when control room operators shifted the configuration of the Unit 1 A train Solid State Protection System (SSPS) while performing refueling outage activities. Results: One violation was identified. The violation involved an inadequate SSPS operating procedure. The procedure did not contain appropriate instructions to operators that Bypass / Permissive light board indications on the main control board may not reflect actual safety injection block conditions when the SSPS is in Test. The failure to incorporate this guidance contributed to a safety injection and an unnecessary challenge to core safety systems. This event is one of several events during the Unit I refueling outage that occurred during increased levels of activity in the control room. In this case, testing on both trains of the SSPS was being performed concurrently. During normal operation testing is limited to one train at a time. 9306020196 930517 gDR ADDCK 05000424 PDR
- . . _ . - ,, i . - % REPORT DETAILS -1. Persons Contacted Licensee Empl6yees J. Beasley, Assistant General Manager Plant Operations W. Burmeister, Manager Engineering Support
- S. Chesnut, Manager Engineering Technical Support
C. Christiansen, SAER Supervisor
- R. Donman, Manager Training and Emergency Preparedness
- W. Dunn Jr., Unit Superintendent, Operations
- C. Eckert, Senior Technical Specialist SAER
- G. Frederick, Manager Maintenance
M. Griffis, Manager Plant Hodifications
- K. Holmes, Manager Operations
D. Huyck, Nuclear Security Manager W. Kitchens, Assistant General Manager Plant Support
- R. LeGrand, Manager Health Physics and Chemist.y
- G. McCarley, ISEG Supervisor
- M. Sheibani, Nuclear Safety and Compliance Supervisor
- W. Shipman, General Manager Nuclear Plant
- C, Stinespring, Manager Administration
- J. Swartzwelder, Manager Outage and Planning
Other licensee employees contacted included technicians, supervisors, engineers, operators, maintenance personnel, quality control inspectors, and office personnel. Oglethorpe Power Company Representative T. Mozingo NRC Resident Inspectors
- B. Bonser
- D. Starkey
P. Balmain
- Attended Exit Interview
An alphabetical list of abbreviations is located in the last paragraph of the inspection report. 2. Event Description At 8:24 a.m., on April 18, 1993, Vogtle Unit I declared and terminated a Hotice of Unusual Event. The NOUE was declared due to an ECCS injection into the reactor vessel. At the time of the event, 7:56 a.m., Unit I was in Mode 5, Cold Shutdown, and the RCS was solid. The plant had been shutdown since March 13 for refueling outage IR4. The RCS had been filled,. vented, and o
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. i ' 2 - l - , RCP sweeps were in progress to complete the fill and vent process. l Train A RHR was in the shutdown cooling mode of operation. RCS pressure was at 345 psig and temperature was at 120 degrees F. The train A CCP ' was in service providing normal charging and seal injection for the ' RCPs. Both SI pumps and both CS pumps were tagged out as required by procedure. Both trains of SSPS were in Test. Train B 120v AC vital
instrument bus IBYlB, which powers channel II input relays to SSPS, was ! deenergized for PM. I&C was also performing A train delta temperature ! , ' channel calibrations as part of the power uprate modification. The Unit I control room was making preparations to perform slave relay testing on AFW to retest some A train ESFAS test valve failures. ' In order to perform the slave relay testing the SSPS had to be returned
l to the Operate mode. An operator was assigned the task of performing i procedure 13503-1, Reactor Control Solid-State Protection System, to return the system to the Operate Mode. This procedure provides instructions on how to energize and de-energize the SSPS. The operator < followed the procedure and ensured that the Low Pressurizer Pressure SI and the Low Steamline Pressure SI were blocked. The SS also verified these blocks by observing the illuminated lights on the BPLP on the main control board. When the operator subsequently took the SSPS out of Test ,' to Operate, an A train SI occurred on Pressurizer Low Pressure and Low Steamline Pressure. !
When the SI occurred, BIT valve IHV-8801A opened, and the A CCP, which was already running and supplying seal injection, . injected into the , core. The A DG started due to the SI signal. Operators immediately l recognized the SI and_ closed valve IHV-8438, a motor operated valve in , the A CCP flow path, which stopped the injection of water into the core ~; and mitigated the pressure transient. CCP A ran a little over 60 seconds in the SI alignment before the normal alignment could be
restored. An SI signal cannot be reset for 60 seconds. RCS pressure peaked at about 450 psig. The RHR suction reliefs (450 psig setpoint), which were providing low temperature overpressure protection, opened and relieved to the pressurizer relief tank. The DG was secured after about , eleven minutes. All operable major ECCS components operated normally during the 1 transient. When valve IHV-8438 was closed, stopping the injection into the core, charging system pressure increased and opened the alternate i miniflow relief valve as designed. One area of concern was identified , when CCP A alternate miniflow valve, IPSV-8501A, bellows ruptured . ' leaking water in to the area outside the CCP A room (see discussion- below). 3. Cause Of The Event Due to IBYlB being deenergized, the bistables associated with channel two of SSPS were in their tripped condition (lights illuminated on the TSLB), including the P-11 pressurizer pressure bistable. The channel
one SSPS P-ll bistable was also tripped due to I&C testing. This i resulted in two of three P-11 pressurizer pressure bistables in a , ,
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. . - , 3 - tripped condition, simulating a high RCS pressure condition (pressure- - > 2000 psig). The P-11 setpoint is reached on decreasing RCS pressure at about 2000 psig. This is a permissive that allows the SI signals for Low Pressurizer Pressure and Low Stear Line Pressure to be blocked during a planned cooldown and depresseization of the RCS. When 2 of-3 pressurizer pressure channels sense less than 2000 psig the P-11 interlock is activated allowing manual block of. pressurizer low pressure SI and low steam line pressure SI. The P-11 permissive activates in a two out of three logic. When the SSPS was taken to Operate, SSPS inputs were reprocessed. As a result of the two bistables simulating a high RCS pressure condition, P-ll unblocked and the SSPS processed the SI signal for low Pressurizer Pressure and Low Steam Line Pressure. Only an A train SI signal was processed because Train B was left in Test. With the SSPS in the Test mode the block status lights on the BPLB, indicating pressurizer low pressure SI and low steam line pressure SI were blocked, had been lit. In addition, the two P-11 bistables on the TSLB were also illuminated indicating that the P-11 permissive was not met. This should have indicated to the operators that the P-ll was not met. The operators, however, verified that the block lights illuminated, as required by the procedure, and did not question the P-11 bistable lights. The inspectors concluded after a review of this event that the system operating procedure, 13503-1, on which the operators had relied to take the SSPS to P erate, had not provided adequate guidance to alert them that the block indications they had observed on the BPLB might not reflect actual block conditions. The operators present, however, were trained on the SSPS and should have recognized the potential for an SI, ' had they also identified and questioned the presence of the P-11 bistable lights. The inspectors also concluded that although the SSPS procedure was inadequate and was a significant contributor to this event; had the SS questioned the wisdom of performing more than one activity on the SSP 5, and limited the activity on this complex circuitry, this would have provided an additional barrier to this type event. The inspectors had no indication throughout this inspection, however, that pressure was being placed on the control room operators to complete these outage activities too rapidly or unsafely. This is identified as Violation 50-424/93-07-01: Inadequate SSPS Operating Procedure Results In Safety Injection. 4. Check Of Alternate Miniflow Valve and RHR Reliefs When the SI occurred CCP A miniflow was realigned to the alternate miniflow line. When the injection into the core was terminated, shortly after the SI signal initiated the event, the alternate miniflow relief valve, IPSV-8510A, lifted to relieve pressure and direct charging flow back to the RWST. When the relief valve lifted, the bellows ruptured, , leaking water into the valve room. The problem with the valve bellows has been recognized by the licensee as a maintenance and a room contamination problem. Previous inspections bf this relief system at Vogtle have not identified the bellows problems or other design problems
_ _ _ _ - _ _ _ _ M . 4 > identified at the Shearon Harris Nuclear Power Plant, which included valve chatter and water hammer (see NRC Inspection Reports 50-424,425/ 92-18, 92-20, 92-27). The Vogtle alternate miniflow design is currently under review by the NRC. Following this event the inspectors walked down the RHR relief valve lines and checked the alternate miniflow line for signs of damage or waterhammer. There was no observable evidence of damage on either system. The inspectors found the bonnet leakoff connection plugged on RHR relief valve IPSV-8708A. With this plug installed, the setpoint of the relief valve can be affected should there be leakage into the valve bonnet. On removal of the plug, however, no bonnet leakage was evident. The alternate miniflow relief valve, IPSV-8510A, was removed and repl aced. The inspectors observed a bench test of the damaged valve. The setpoint of the relief valve was found to be at 2450 psi. The setpoint pressure on these valves is 2200 1 66 psi. A licensee review of past operating experience with these valves indicated that the setpoints can experience drift. However, no instances were identified where the drift would have prevented a valve from performing its intended safety function. The licensee has committed to verify the setpoint of one of these valves at each refueling outage. The licensee is also considering several design changes to the alternate miniflow system. 5. Exit Meeting The inspection scope and findings were summarized on May 7, 1993, with those persons indicated in paragraph 1. The inspector described the areas inspected and discussed in detail the inspection findings listed below. No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during the inspection. Item No. Description and Reference VIO 424/93-07-01 Inadequate SSPS Operating Procedure Results In Safety Injection (paragraph 3) 6. Abbreviations AFW - Auxiliary Feedwater System BIT - Boron Injection Tank BPLB - Bypass / Permissive Light Board CCP - Centrifugal Charging Pump CFR - Code of Federal Regulations CR - Control Room CS - Containment Spray DG - Diesel Generator i ECCS - Emergency Core Cooling Systems ESFAS - Engineered Safety Feature Actuation System I&C - Instrumentation and Controls IR - Inspection Report L . . .
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- 5 ISEG - Independent Safety Engineering Group LCO - Limiting Condition for Operation LER - Licensee Event Report MOV - Motor Operated Valve MWO - Maintenance Work Order NOUE - Notice Of Unusual Event NPF - Nuclear Power Facility NRC - Nuclear Regulatory Commission PM - Preventive Maintenance psig - Pounds Per Square Inch RCP - Reactor Coolant Pump RCS - Reactor Coolant System RHR - Residual Heat Removal R0 - Reactor Operator RWST - Refueling Water Storage Tank SAER - Safety Audit And Engineering Review SI - Safety Injection SS - Shift Supervisor SSPS - Solid State Protection System TS - Technical Specifications TSLB - Trip Status Light Board USS - Unit Shift Supervisor VIO - Violation IR4 - Unit 1 Refueling Outage Cycle Four 4 l . . _ _ - __ __- .. -- }}