IR 05000269/1980023
| ML19337A888 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/28/1980 |
| From: | Jape F, Martin R, William Orders NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19337A880 | List: |
| References | |
| 50-269-80-23, 50-270-80-20, 50-287-80-17, NUDOCS 8009300640 | |
| Download: ML19337A888 (13) | |
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G UNITED STATES
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8 " ) 0 (f' ' n b NUCLEAR REGULATORY COMMISSION J-E REGION 11 y,
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Report Nos. 50-269/80-23, 50-270/80-20 and 50-287/80-17 Licensee: Duke Power Company 422 South Church Street
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Charlotte, NC 28242 Facility: Oconee Nuclear Station Docket Nos. 50-269, 50-270 and 50-287 License Nos. DPR-48, DPR-47 and DPR-55 Inspection at Oconee Nuclear Station near Seneca, South Carolina Inspectors:
'741Mw ds 7/2#/fD F. Jape g
Date ' Signed A4a h 7b f/S*1'
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Date Signed
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D. Myers tf D(te Signed
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Approved by:
R. D. Martin, Section Chief, RONS Branch Dafe Si(gned
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SUMMARY Inspection on June 2-30, 1980 Areas Inspected This routine inspection involved 225 resident inspector-hours on site in the areas of test witnessing, direct verification of system lineup, followup of NUREG-0578 items, followup of IEB 79-27, maintenance activities and plant operations.
Results Of the six areas inspected, no items of noncompliance or deviations were identified in five areas; one item of noncompliance was found in one area (Infraction:
using a Unit 1 procedure to perform a periodic test on Unit 2 Paragraph 8.f.).
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DETAILS 1.
Persons Contacted
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J Licensee Employees
- J. E. Smith, Station Manager
- J. M. Davis, Superintendent of Maintenance
- J. N. Pope, Superintendent of Operations
- T. B. Owen, Superintendent of Technical Services
- R. T. Bond, Licensing and Projects Engineer
- J. Brackett, Senior QA Engineer Other licensee employees contacted included 20 operations supervisors, 8 technicians, 25 operators, 6 mechanics, station health physicist, 3 mainte-nance coordinators and 6 maintenance supervisors.
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- Attended exit interviews 2.
Exit Interviews The inspection scope and findings were summarized on June 6, 13, 16, 20 and 30, 1980, with those persons indicated in Paragraph I above. The inspection findings were acknowledged without significant comment. Licensee management assigned immediate attention to the noncompliance item identified (paragraph 8.f.) and to resolution of the administrative items in NUREG-0578, paragraph 5.
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Licensee Action on Previous Inspection Findings Not inspected.
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Unresolved Items Unresolved items were not identified during this inspection.
5.
Implementation of Short Term Lessons Learned, NUREG-0578 NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations," and NRC letter to all operating nuclear power plants, dated September 13, 1979, established requirements for all operating sites, including Oconee Nuclear Station. Implementation was reviewedJat the site on Jan. 14 and 15, 1980, by NaC staff. The status and findings were reported by letter dated April 7, 1980, to Duke Power Company.
Items requiring additional followup were completed by the Resident Inspector and are reported below. The numbered designation of each item is consistent with that used in NUREG-0578.
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2.1.3.a Direct Indication of PORV and Safety Valve Position Nuclear Station Modification 1391 has been completed on all three Oconee units.
The reactor coolant system relief (PORV) and code safety valves have been provided with acoustical accelerometers to' provide a direct position indication of thae valves. The inspector verified the physical installation of the monitors and witnessed the functional test for all three units.
(See IE Reports 50-269/80-7, 50-287/80-11 and 50-270/80- for details on testing.) Also, records for installation and component specifica-tions were examined. There were no questions concerning the installation and functional testing for this item.
2.1.3.b Instrumentation for Detection of Inadequate Core Cooling
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Nuclear Station Modification 7441 "T Sat Meter," has been installed and is working on all three Oconee units. The on-line computer calculates and monitors the margin to saturation for each reactor coolant loop. When the reactor coolant system is approaching saturation the program gives an annunciator alarm, a video (computer) alarm, and the indication meter turns
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red.
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During performance of TI/3/B/320/5, "ICS Loss of Power Test", on April 16,
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1980, a problem with the T Sat meter was identified. A programming change
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was made on all three Oconee units to correct the problp. The T Sat meter
program ignores the KI inverter powered signals if they indicate 350 i 10*F while the redundant KU inverter powered signals do not. Thus the T Sat meter indication will be maintained on loss of ICS power. The use of the T
Gat meter has been incorporated in EP/0/A/1800/30, " Inadequate Core Cooling".
The inspector had no questions or concerns following review of this procedure l
and discussions with operations and performance personnel on the use and adequacy of the T Sat meter.
2.1.4 Containment Isolation Provicions Nuclear station modification 1392 has been completed on all three Oconee units. This modification provides diverse parameters to initiate contain-ment isolation and prevents - automatic opening of any isolation valve upon reset.
Reopening of containment isolation requires deliberate operator actions.
The installation has been functionally tested on all three Oconee units and the following procedures have been revised to accommodate this modification.
1P/0/A/310/7A, HPI and RB Isolation Ch-1 Functional Test
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1P/0/A/310/8A, Logic Subsystem 2 HPI & RB Isolation Ch-2 Functional
_3t IP/0/A/310/7C, RB Isolation & Cooling Ch-5 Functional Test
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IP/0/A/310/8C, RB Isolation & Cooling Ch-6 Functional Test IP/0/A/310/12A, HPI S.RB Isolation Ch-1 On-Line Test 1P/0/A/310/13A, HPI & RB Isolation Ch-2 On-Line Test IP/0/A/310/12C, RB Isolation & Cooliag Ch-5 On-Line Test
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IP/0/A/310/13C, RB Isolation & Cooling Ch-1 On-Line Test PT/0/A/160/3, RB Coolers ES and Performance Test PT/0/A/160/12, HPI System ES Test
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The inspector had no questions or comments on this item.
2.1.5.c Operating H Control System
The licensee has provided procedure HP/0/B/1006/05, Procedure for Quantifying High Level Gaseous Radioactivity Releases During Accident Conditions. This procedure was reviewed by the inspector and was found to be adequate for the intended purpose.
2.1.6.1 System Integrity A leak reduction program has been implemented on all three Oconee units.
Leakage measurements have been taken and a leak reduction program is underway.
Systems will be checked periodically for leakage.
The program includes the following procedures:
PT/0/A/150/22, CS Stream Leakage PY/0/A/150/23, HPI System Leakage PT/0/A/150/24, RB Spray Steam Leakage PT/0/A/150/25, LWD System Leakage PT/0/A/150/26, GWD System Leakage PT/0/A/251/13, CC Check Valve Functional Test PT/0/A/251/14, FW Check Valve Functional Test PT/0/A/251/15, BS Check Valve Functional Test PT/1/A/251/11, Quench Tank Check Valve Functional Test PT/2/A/251/11, Quench Tank Check Valve Functional Test PT/3/A/251/11, Quench Tank Check Valve Functional Test PT/1/A/251/12, LPI Check Valve Operability Test PT/2/A/251/12, LPI Check Valve Operability Test PT/3/A/251/12, LPI Check Valve Operability Test The inspector reviewed the above procedures and found them adequate for the intended purpose.
2.1.8.a Post-Accident Sampling Procedures, listed below, are in place and ready for use for taking reactor coolant samples and containment atmosphere samples during an accident.
These procedures were examined by the inspector. There were no comments or
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questions.
CP/0/A/200/8, " Chemistry Sampling System"
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j HP/0/B/1006/5, " Procedure for Quantifying High Level Gaseous Radioactivity l
Releases During Accident Conditions."
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2.1.8.b High Range Effluent Radiation Monitors
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The inspector reviewed the licensee's procedure for quantifying high level
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radioiodine and particulate effluents from the plant.
The procedure, HP/0/B/1006/05, was found to be adequate for the intende/ iunction.
2.1.8.c In-Plant Iodine Instrumentation
The inspector reviewed procedure HP/0/B/1006/05 which provides for accurately determining airborne iodine concentration throughout the plant under accident
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conditions. The procedure adequately covers the requirement.
2.2.1.a Shif t Supervisor Responsibilities The authority and responsibility in an emergency situation o'f the shif t supervisor on duty has been expanded and is addressed in Administrative Policy Manual Section 3.1.2.4.
In addition the Position Guide for the Shift Supervisor and Assistant Shift Supervisor have been examined and found to limit duties and responsibilities so that the supervisors will not be distracted by administrative assignments.
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2.2.1.b Shift Technical Advisor I
I Since January 1, 1980, a Shif t Technical Adviso; (STA) has been assigned
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each shift. His duties and responsibilities are described in an intra-station letter.
In addition, Station Directive 4.2.7 describes the Safety Review Engineer (SRE) functions.
The STA and the SRE responsibilities
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complement; one another.
The inspector interviewed a number of STA's and SRE's and examined their
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actions.
The program established for all three Oconee units appears to satisfy the requirements of the item.
2.2.1.C Shift and Relief Turnover Procedures Station Directive 3.1.8 describes the policies for shift relief and turnover.
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A separate turnover checklist is provided for the shift supervisors, control
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operators and the nuclear equipment operators. During the past several months, completed forms have been examined to determine if the plant status i
is accurately reflected. In addition, interviews with all levels of plant personnel have been held to determine if the employees utilize these data.
The inspector found the records to accurately reflect plant status and that personnel are using the turnover sheets as intended. The inspector finds the system to be an effective technique for adequate shift relief and turnover.
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2.2.2.a Control Room Access Station Directive 3.1.31 delineates the Oconee Nuclear Station policy for access tb' the control rooms. Responsibilitiy and authority are designatad within this instruction. In addition, access points to each control room
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are posted. The inspector has observed effective implementation of this directive during testing and minor crises when an overload of personnel congregate within the control area. The inspector finds this instruction and its implementation to be effective and satisfactory, 2.2.2b. On-Site Technical Support Center Station Directive 3.8.5, Appendix D describes the establishment of an on-site center for use during site emergencies. The computer rooms are designated as the space to be set up for this purpose. Communications are available with the Control Rooms and plant parameters can be monitored by CRT display without entry into the control rooms or interference with activities in the control rooms.
The inspector finds the directive to adequately cover the requirement.
2.2.2.c Operational Support Center (OSC)
Station Directive 3.1.31 " Control Room Access and Authority" states that the Nuclear Equipment Operators shall assemble in the kitchen areas to be
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available for assignment. This directive adequately covers this item.
6.
IEB 79-27 " Loss of Non-Class I-E Instrumentation and Control Power System Bus During Operation."
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IEB 79-27 was issued November 30, 1979, following an event that occurred at i
Oconee Unic 3.
This is partially described in R0-287/79-13.
The IEB requested information regarding three areas. Followup action by the resident inspector is concentrated on Item 2 of the Bulletin.
Item 2 required a review and revision of existing procedures, or preparation of a new emergency procedure, to cover the event. DPC has provided EP/0/A/
1800/31, " Loss of KI Bus (and Control Room Indication Powered from KI)."
This procedure was reviewed by the inspector and found to cover the event.
In addition, Nuclear Station Modification 1446 has been installed on all three Oconee units. This modification provided a redundant transfer switch that will automatically transfer power supply to the XV inverter upon loss of the KI inverter. Secondly, Nuclear Station Modification 1531 has been installed and tested on Oconee Units 2 and 3.
Installation on Oconee Unit 1 is scheduled to be completed and tested following the outage beginning no later than July 16, 1980. The purpose of NSM 1531 is to provide a redundant source of power from the KV inverter to selected indications and controls normally powered from the KI inverter.
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The emergency procedure, EP/0/A/1800/31, has been revised to reflect the two modifications. The licensee's response dated March 7, 1980, to Item 2 of the IEB is considered responsive and adequate. Items 1 and 3 of the IEB l
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7.
Loss of Control of Resin During Transfer During routine transfer of spent resin from Unit I the spent resin storage tank (SRST) was overftlled causing resin to be spread to the vent header system. The cause appears to have been ineffective level monitoring of the SRST. Resulting high radiation areas (approximately 40 R/br were promptly located and designated in auxiliary building. Clean-up procedures consisting of header flushing have been progressing slowly due to rough internal surfaces of the carbon steel vent header retaining resin beads. New resin has been loaded and the affected demineralizer returned to service. To prevent recurrence of this event the licensee plans to install a screen in the overflow line similar to those in Units 2 and 3.
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Test Witnessing Unit 2 The following test activities were witnessed by the inspector to ascertain crew performance, conformance with license and procedural requirements and adequacy of test results.
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Reactor Coolant System Leak Test a.
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The reactor coolant system leak test, PT/0/A/200/46 is performed to assure reactor coolant system integrity prior to re turn to operation following opening, modification or repair of the reactor coolant system.
The reactor coolant system pressure is increased to 2285 psig and a visual inspection is made to verify that no reactor coolant leakage exists.
Any other leakage to the reactor building atmosphere is evaluated prior to resumption of reactor operation.
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The inspector reviewed procedure PT/0/A/200/46 which appeared to be technically adequate and in compliance with Technical Specification
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requirements.
The inspecter witnessed the performance of PT/0/A/200/46 on June 20, 1980, and observed that the test appeared to be performed as required by the approved procedure.
During the inspection, several inches of water were discovered in the area under the reactor vessel. Investigation by the licensee revealed that the drain from this area had been plugged, and the water apparently collected from around the seal plate during refueling. The plugged line has been opened and the water drained.
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Test results met the acceptance criteria.
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PORV and Code Safety Valve Position Indication Reactor coolant system relief (PORV) and code safety valves have been provided with a positive position indication in the control room. The
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system was installed as described in Nuclear Station Modification
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1391.
Position of the three relief valves, RC-66, RC-67 and RC-68, is monitored by the TEC valve monitoring system. Accoustical-accelerations propor-tional to valve position are generated as flow is established through the discharge piping of the valve. These signals are detected by an accelerometer strapped to the piping and are converted to a voltage signal. The voltage is proportional to position and is displayed in the coctrol room. An alarm module initiates a Statalarm whenever flow rate indicates the valve to be open greater than 25%. Test procedure, TT/2/B/1391/0, " Pressurizer Valve Monitor Calibration" was provided for initial checkout and calibration of the system.
-Performance of this test was witnessed by the inspector. The test was conducted at hot shutdown conditions.
The modification performed adequately and met acceptance criteria.
In addition, the following procedures were reviesed to incorporate the modifications:
EP/0/A/1800/1, " Load Rejection"
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EP/0/A/1800/2, " Turbine Trip" EP/0/A/1800/3, " Reactor Trip" These were reviewed by the inspector and found to be adequate.
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High Pressure Injection System Modification (NSM 1080)
Previous small break loss of coolant analyses considered the reactor
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coolant pump (RCP) suction line as the limiting break location for small breaks. Assuming only one of two trains of the high pressure injection (HPI) system were available, the installed system was adequate to provide the necessary core cooling. However, it has been determined that the limiting break location for small breaks is the pump discharge of the reactor coolant system cold legs and not the RCP suction. A revised safety analysis performed by the NSSS vendor, for breaks at
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this location, revealed that one train of HPI flow was insufficient to l
maintain core coverage.
Interim measures have been in effect that require operator actions outside the control room. Permanent plant
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modifications have been installed on Ocenee Unit 1.
The modifications consist of a cross-connect line between the A and B HPI discharge lines down stream of the Engineered Safeguards (DS) valves and another tie-line connecting this cross-connect Hae and the B and C HPI pumps discharge header. The newly instalP 4 1alve= are manually controlled, electrically operated valves, eq & < c ' being manipulated from the control room. Thus, operator ay. ;;p, ; side the control ioom will be eliminated.
The inspector reviewed and witnessed the perforgance of the functional test; TT/2/A/203/11, verifying operability of the station modification.
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The acceptance criteria requires a flow rate via the cross-connect lines of greater than or equal to 450 gpm with 'RCS pressure at 600 psig. The system performed adequately as expected and was declared operable.
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d.
Integrat md Control System Loss of Power Test As a result of experience gained from the Crystal River-3 incident of February 26, 1980, a modification to the power supply for the non-nuclear instrumentation (NNI) was designed by DPC and has been installed on Oconee Unit 2.
The modification is described in Nuclear Station Modification 1531, (NSM 1531). This NSM provides a redundant source of power to all indications and control loops in the Integrated Control System, (ICS), necessary to reach and maintain the reactor at hot h+ shutdown upon loss of the normal power supply to the NNI. The indications and controls needed to maintain the plant at hot shutdown are stated in an April 1, 1980, letter from DPC to NRC. NSM 1531 incorporated these indications and controls.
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Following the installation of NSM 1531, a series of tests were performed to verify system performance. This test is described in TI/2/B/320/05,
" Integrated Control System Loss Power Test".
An NRC order confirming
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DPC's commitment to perform such a test was issued April 17, 1980.
The test was conducted on June 6,1980.
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Prior to running the test, comments on the test procedure were discussed with licensee representatives and resolved. The test was witnessed by the inspector.
Several discrepancies identified during the testing have been reviewed by the licensee and either resolved or corrected and retested.
The inspector concurred with the resolution in each Case.
Results of the testing have been incorporated into Emergency Procedure EP/2/A/1800/31, " Loss of 3KI Bus (And Control Room Indications Powered From 3KI)".
The inspector discussed the NSM and the testing with several operators and supervisors to verify that they were acquainted with the change. Personnel were found to be familiar with the modifi-cations and !ad received training through the: requalification training program on this change.
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Steam Generator Automatic Level Control An automatic steam generator level control system has been installed.
Details are described in Nuclear Station Modification 1275, Part K;
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The system consists of two redundant trains of level contr'l for each, o
steam generator. Whenever there is an emergency start of the motor driven emergency feedwater pumps, the level control system will throttle emergency header valves, FDW-315 and FDW-316, to control the level at
25 inches. If there is a loss of all four reactor coolant pumps, the level will be controlled at 240 inches.
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-9-Test procedure, TT/2/A/275/5W, " Steam Generator Auxiliary Level Control System Tuning Procedure", was performed with the plant at hot shutdown and at 15% power.
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The inspector witnessed the test performance at l5% power and reviewed the test results of both tests. The system is considered operational and the testing successfully completed.
The level control system is totally independent of the Integrated Control System, and controls level as expected.
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Emergency Feedwater Pump Test On June 11, 1980, the inspector witnessed testing of the Oconee Unit 2 Turbine Driven Emergency Feedwater Pump Overspeed Test. The inspector detected that the procedure being used at the test site, the " Working Copy" of the procedure, was an Oconee Unit 1 procedure, OP/1/A/1106/06.
The inspector also noted that the procedure had not been signed off as having been checked against the Controlling Copy of the procedure for validity-and prevalence. The inspector discussed the procedural error
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with the test coordinator. The correct, approved procedure was obtained prior to continuing the test on June 12.
While witnessing the testing on June 12, the inspector detected that the actual valve line-up did not agree with the requirements of the procedure.
For example, valve MS-90 (Main Steam Supply) was found locked open when the procedure required that the valve to be closed and Red Tagged.
Step (4.22) in the procedure requires that MS-90 be closed. This step was not signed-off as being completed on the " working copy" and review of the " control copy" revealed the step had been deleted by the operator.
Oconee Station Directive 4.2.1 allows limited deviation from written approved procedures, but in this case deletion of portions of the test procedure, including step 4.2.2, did not abide with directive require-ments.
Using the incorrect procedure when performing the test on June 11, 1980, and failure to follow the requirements for changing test procedure on June 12, 1980, is considered to be noncompliance with Technical Specification 6.4.1 which requires that the station be operated and maintair.ed in accordance with current, written approved procedures.
The two above examples of the licensee's failure to follow procedures constitutes an infraction. (270/80-20-01).
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Emergency Power System Testing The'following tests were witnessed by the inspector to verify confor-mance to license and procedural requirements for surveillance testing:
(1) PT/2/A/610/1H Emergency Power Swi'tching Logic (EPSL) Standby i
Breaker Closure CN. A & B
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(2) PT/2/A/610/1J EPSL Actuation Keowee Emergency Start Test (3) PT/0/A/202/12 High Pressure Injection System Engineered Safeguards Test
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Items Inspected Included:
(1) Test personnel were using the appropriate revision of the procedure (2) Test prerequisites and initial conditions of procedure were met (3) Tests were performed and required changes were made in accordance with licensee procedures (4) Test data and component response were reviewed by test supervisors to assure adequacy of results (5) Proper coordination between operations and testing personnel were maintained.
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Items inspected were satisfactory; however, numerous administrative and minor procedure deficiencies were noted.
Examples are:
the racking out of an energized 4160V breaker, referencing non-existant breakers numbers, unexpected loss of main feeder bus voltage indication with the bus energized. While significant test delays resulted, no personnel or equipment damage was noted. The inspector observed in the exit interview that a trend toward insufficient test review may be I
developing.
During the HPI ES Test (PT/0/A/202/12), the inspector noted that when manual control of ES components was taken from the ES panel after a simulated ES signal,
"C" HPI pump tripped to the "off" position unexpectedly.
No other ES components changed state.
Subsequent review of system logic diagrams indicated that the pump responsed as designed. The pump must have its controlle.r at the unit control board in the " closed" position to prevent being' tripped from the ES panel.
Normal position of the controller is "off".
This condition was not addressed in the test procedure and further investigation revealed it
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was not addressed in the HPI operating or emergency procedures. The licensee was requested to review pertinent procedures and make changes if necessary to ensure that operations personnel would be alerted to j
the pump response.
i P1/2/A/610/13 was a new test developed to meet present license Amendment 82.
The test appeared to be in full compliance with requirements of the Amendment.
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Correcti.ve Maintenance During the reactor system leak test at hot shutdown conditions, a small
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leak was observed at the flange of a safety relief valve, RC-68.
The
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i licensee's plans to repair this were examined and the repair work was
witnessed by the inspector.
The job required a modification, which was prepared in accordance with Station Directive 4.4.2, and a procedure, which was prepared as described in Section 3.3.3 of the Administrative Policy Manual. These documents were reviewed by the inspector and found to be adequate.
The repair was to use a Furmanite process to seal the leaking flange without
replacing the gasket. An evaluation of their procedure was performed by i
the licensee and included within the modification plan.
The chemical content of the Furmanite material to be used was also included within the modification plan.
l The inspector found that this activity conformed with applicable regulatory
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guides and Technical Specifications. The documents reviewed were as follows:
NSM-1594, Furmanite Inlet Flange to RC-68 i
TN/2/A/1594/0, Procedure for Implementation and Verification of NSM
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-0N-1594 10.
Broken Holddown Springs, R0-269/80-15 In response to the May 16, 1980, Babcock and Wilcox notification on the possibility of defective fuel assembly holddown springs licensee investiga-d tion of spent fuel assemblies on site was made.
LER 50-269/80-15 -was i
submitted to document discovery of four broken holddown springs and described
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corrective actions program.
Implementation of corrective actions has been verified by the inspector.
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11. Fire Detection Capability On or about June 2,1980, while performing maintenance, the licensee detected that the control room fire alarm statalarms will not reannunciate or reflash on a redundant detector trip in the Honeywell fire detection systems.
This failure to reannunciate effectively renders portions of the station without adequate fire detection capability during periods when one or more detectors are in the alara state or have failed.
The licensee is preparing a station modification which will provide the necessary annunciation capability.
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12. Systems Alignment Verification Following the Oconee Unit 2 refueling outage and immediately preceeding Reactor Building close-out, the inspector performed a walkdown of the accessible portions both inside and outside the Reactor Building, of the
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systems listed below, the walkdown was performed to ascertain that the licensee's systems lineups reflect their respective lineup procedures and drawings, to identify equipment conditions, and to identify items which might degrade performance, such as hangars and supports, and to observe accessible piping for leakage or leakage paths.
System Procedure Drawings High Pressure Injection OP/2/A/1104/02 PO-101A-2 PO-101B-2 Low Pressure Injection OP/2/A/1104/04 P0-102A-2 Reactor Building Spray OP/2/A/1104/05 PO-103A-2 Component Cooling OP/2/A/1104/08 P0-144A-2 Reactor Coolant OP/2/A/1103/02 PO-100A-2 Core Flood OP/2/A/1104/01 PO-102A-2 Within the areas inspected, no apparent items of noncompliance or deviations were identified.
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