IR 05000400/1987002
| ML20207T559 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/19/1987 |
| From: | Burnett P, Jape F, Mathis J, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207T550 | List: |
| References | |
| 50-400-87-02, 50-400-87-2, NUDOCS 8703240129 | |
| Download: ML20207T559 (10) | |
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ATLANTA, GEORGI A 30323 L<.
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_ Report No.: 50-400/87-02 Licensee:. Carolina Power and Light Company P. O. Box 1551
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Raleigh, NC 27602 iDocket Nos.: 50-400 License Nos.: NPF-63
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~ Facility Name: Harris 1
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Inspection Con cted: January 7-29, 1987 Inspectors:
'.[ Naikk 2 fl4[67 P. T. Bbrnett Dat'e Signed D. f. hAAL, 2 /19 /8 7 J.8
. Mathis Dat'e Signed h
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Date Signed Accompanying Personnel:
J. MacDonald K. VanDyne R. H. Bernhard Approved by:
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F. Jape, Se'ction Chief, TPS(/
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Date Signed Engineering Branch Division of Reactor Safety SUMMARY
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Scope:.This routine, unannounced inspection was conducted in the areas of start-up test evaluation and review, witnessing of the loss of offsite power and shutdown from-outside the control room tests, and review of surveillance procedures.
Results: No violations or deviations were identified.
G703240129 870220 PDR ADOCK 05000400
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- R. A. Watson, Vice President, Harris Project
- J. L. Willis, Plant General Manager
- G. C. Campbell, Manager, Maintenance
- R. J. Duncan, Test Program Development Engineer, Technical Support
- J. L. Harness, Assistant Plant General Manager, Operations C. S. Hinnant, Manager-Startup W. M. Peavyhouse, Scheduling Coordinator, Technical Support
- J. P. Thompson, Operations Supervisor
- D. L. Tibbets, Director, Regulatory Compliance
- R. B. Van Metre, Manager, Technical Support
- W. R. Wilson, Principal Engineer, Technical Support
- R. R. Wojnarowski, Reactor Engineering Leader, Technical Support Other licensee employees contacted included engineers, technicians, operators, mechanics, security office members, and office personnel.
NRC Resident Inspectors
- G. F. Maxwell S. P. Burris
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on January 29, 1987, 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 licensee made a commitment to resolve the following item before exceeding 30% power:
Inspector Followup Item 400/87-02-01:
Resolve Overspeed Trips of the Turbine Driven Auxiliary Feedwater Pump paragraph 7.
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 This subject was not addressed in the inspectio e
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. Unresolved Items Unresolved items were not identified during'this inspection.
5.
Low Power Physics Testing a.
Deferred Precritical Testing (72596)
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The - following completed test procedures were reviewed to assure that the' acceptance criteria were satisfied. and all test deficiencies resolved:
(1) 9102-S-06, Pressurizer Spray,- Heater Capability and Continuous Spray -Flow Setting, was performed on January 1-2 and January 5, 1987.
There were six _ test exceptions written.
During the performance.: of the pressurizer heater capability test, the pressurizer pressure response did not fall within the acceptance-
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criteria shown on Figure 2 of CQL-SU-2.1.5. The plant measured a pressure that was lower than the minimum expected pressure. by 4.0 psi at 240 seconds after the transient initiation. The degree of deviation noted in the pressure response is acceptable to Westinghouse. This is based on analysis that have been performed which indicate significant pressurizer heat capacity reduction do not materially affect the plant. transient capability. Based on Westinghouse analyses test result was accepted "as is" by the licensee.
(2) 9102-S-20, Incore Movable Detector System. This test demonstrated that the Incore Nuclear Instrumentation System can remotely position the incore neutron detectors for the purpose of core flux mapping and to supply the appropriate digital and analog signals to the plant computers. All test exceptions appeared reasonable and did not invalidate the test.
(3) 9102-S-04, Reactor Coolant System Flow coast-down was performed on January 8, 1987.
There was one test exception written against this procedure. Flow coast-down upon loss of flow from all three reactor coolant pumps were measured and insured that the flow coast-down time constant is more than design values. Also this test verified that the time delay associated with reactor protec-tion system actuation for a loss of flow accident is less than design values.
The Reactor Coolant low flow trip loop time response determined during this test equalled 0.73 sec. The Flow coast-down time constant determined was 13.13 seconds, b.
Natural Circulation (72576)
Natural Circulation Test,1-9103-S-23 was performed in accordance with the procedure. The inspector verified that the procedure, 1-9103-S-23 was approved and any changes to the procedure did not invalidate the
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intention of' the procedure.
This test verified natural-circulation heat removal capability of the Reactor Coolant System can be maintained with a loss of pressurizer heaters. Natural circulation was maintained for 30 minutes with pressurizer pressure remaining above 2000. psi during the test.
Data obtained during stable natural circulation operations is comparable and within specified conditions to those values obtained at a prototype design plant (North Anna) for which equivalent tests have been conducted.
The AT acceptance criteria was satisfied and additional operator training is not required to close TMI Action Item 1.G.1 during normal requalification periods.
See para-graph 10 for more details on TMI action item 1.G.I.
c.
Flux Distribution (72572)
Test procedure 9103-5-13 (Revision 1), Flux Distribution Test (ARD, HZP), was performed on January 7,1987, in conjunction with :.urveil-lance' test OST-10 (Revision 3), Hot Channel Factors.
The indicated power level on the power range nuclear instruments was less than 5%,
and that indicated by through-the-core change in reactor coolant temperature was.1.5%.
Inspite of the low power level, good agreement was obtained between the normalized predicted and measured power distributions. The. heat flux hot channel factor (Technical Specifica-tion 3.2.2), the nuclear enthalpy rise hot channel factor (Technical Specification 3.2.3), and the quadrant power tilt ratio (Technical Specification 3.2.4) were all acceptable for power operation up through 50% of rated thermal power.
d.
Boron Reactivity Worth (72572)
The reactivity worth of boron was determined from data collected during the measurement of control rod worths.
The measured value was-10.3 pcm/ppmB which is within the range of -16 to -7 pcm/ppmB specified in FSAR Section 14.2.12.2.13.
This result is documented in CP&L Memorandum to File SHF/10-16050 and Westinghouse Memorandum
" Preliminary Report on Zero Power Physics Tests at Shearon Harris Nuclear Power Plant," dated January 6, 1987.
6.
Witnessing Of The Shutdown From Outside The Control Room Test (72583)
This test was performed on January 25, 1987, under the guidance of startup test procedure 9104-S-19 (Revision 1), Remote Shutdown Test.
The control room was abandoned by the minimum shift crew of six people, not including the shift technical advisor, with the reactor operating at about 14% power and under the observation of a backup crew.
The reactor was scrammed by tripping the motor generators for the control rod drives. All subsequent actioas were performed at the auxiliary control panel (ACP) or by radio dirsction from the ACP to two auxiliary operators, part of the minimum shift crew, working out in the plant.
The minimum shift crew performed their
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-4 activities in accordance with EPT-039T (Revision 0), Remote Shutdown Test,
. hich was a ' variation of A0P-004 (Revision 4), Safe Shutdown in. Case of w
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The variation was required in lieu of the abnormal operating procedure because of test requirements to perform no shutdown activities prior to leaving the control room and to leave the reactor coolant pumps running to simulate a source of decay heat.
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The minimum crew performed their duties very well and overcame. with
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equanimity minor. operational difficulties such as the failure of a transfer-of-control indicating relay, a few burned out indicating lights,
.and the tripping on overspeed of turbine driven auxiliary feedwater pump when first started.
All major test acceptance criteria appeared to be satisfied. Following the trip, the unit was stabilized at hot standby for 30 minutes.
It was then cooled down at a rate of less than 100 degree F in any one hour, to the temperature and pressure conditions sufficient to place the residual heat removal system (PHR) in service. The plant was then cooled down 50 degrees ( F) by use of the RHR.
The inspectors monitored the activities of the control room crew to assure their activities were limited to protection of non-essential equipment and monitoring the reactor coolant pumps. There was no direct communication between the control room and the minimum crew at the ACP. The inspectors witnessed all of the activities at the ACP and some of the remote activities throughout the approximately eight-hour test duration.
Review of the licensee's completed evaluation of the test will be performed during a later inspection.
7.
Loss of Offsite Power Test (72582)
The inspectors witnessed licensee performance of test procedure 9104-S-15, Station Electrical Blackout, on January 29, 1987. This included attending the pre-test briefing where personnel involved in the test discussed the test sequence and were given final instructions.
The test began at approximately 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br />. The inspectors observed test activities in the main control room (MCR), auxiliary control panel (ACP), and other locations within the plant. There was adequate coordination among personnel involved in the test, and their actions appeared to be correct and timely during performance of the test.
Except for the items discussed below, the plant response to the test appeared to be satisfactory.
The emergency de lights at the ACP did not come on as required when ac
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lighting was lost during the time period between the loss of offsite power (LOSP) and the restoration of emergency ac power by the emergency diesel generators.
During subsequent troubleshooting the licensee found a loose wire which caused an open circuit in the de system supplying the ACP.
The loose wire was fixed and the emergency de lights at the ACP worked properly.
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. After an automatic start, the turbine driven auxi_11ary feedwater pump.
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(TDAFWP) tripped on an apparent electrical overspeed signal.
The TDAFWP also tripped on an apparent electrical overspeed signal after being manually started during performance of the loss of control room test on January 25, 1987 (paragraph' 6).
During discussion of. this matter licensee personnel - speculated that the overspeed trip which occurred during the loss of control room test may have been due to the pump being. manually started at minimum load conditions, whereas it would normally be aligned for an automatic start at maximum load conditions. The licensee did not know why the pump tripped on over-speed after the automatic start during the LOSP test. The licensee was still investigating both trips at the conclusion of this inspection.
The inspectors stated that the problem with the pump tripping on overspeed should be resolved prior to exceeding 30*4 power. Resolution of this concern will be reviewed during a future inspection and will be tracked as inspector follow-up item 400/87-02-01; Resolution of Turbine Driven Auxiliary Feedwater Pump Overspeed Trips.
During the exit interview the licensee committed to resolve - this iten prior to exceeding 30% power.
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The inspectors will review the approved test results package during a future inspection after the licensee has resolved all test discrepancies and given final approval of the test results.
8.
Reactor Coolant System Leakage Measurements (61728)
By reference to the FSAR and plant drawings, the inspector calculated the plant-specific parameters to be used with the microcomputer program RCSLK9 for independent evaluation of reactor coolant system leakage.
The program is described in NUREG-1107, RCSLK9:
Reactor Coolant System Leakrate Determination for PWRs, December 1984.
The parameter list is given in Attachment I and 2.
Using data provided by the licensee from a completed copy of OST-1026 (Revision 3), RCSLK9 calculated an unidentified leakage rate of 0.65 gpm.
The licensee's result of 0.62 gpm was in acceptable agreement, as were the individual calculations of changes in inventory of the pressurizer, volume control tank, and the balance of the reactor coolant system.
This inspection activity was performed in advance of the licensee's formal response to violation 400/86-96 for an inadequate procedure in this surveillance area.
9.
Thermal Power Measurement (61706)
Using information from the FSAR, plant drawings, and Westinghouse Technical Manual 1440-C264 (12/75), Steam Generator Instructions for Carolina Power and Light Company Shearon Harris Plant Unit 1, the inspector calculated the plant specific parameters necessary to perform independent calculations of
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plant thermal pcwer using the microcomputer program TPDWR2. This program is described in NUREG-1167, TPDWR2: Thermal Power Determination for Westing-house Reactors, Version 2.
Surveillance. Procedure OST-1004 (Revision 2), Power Range Heat Balance, was reviewed and discussed with the licensee.
Once heat balance data are obtained, the results of the procedure will be compared with those from TPDWR2 for acceptability.
10.
Followup of Regional Request (92705)
The inspectors held discussions with licensee representatives concerning the reactor trip that occurred on January 21, 1987. The plant was operating at approximately 10% power (decreasing) when the trip occurred from a high neutron flux signal on one of the intermediate range (IR) channels. Prior to the trip, power was being decreased from approximately 30% to below 10% in order to perform the turbine overspeed trip test.
The licensee attributed the trip to inaccuracies associated with initial adjustments of the IR channels and a conservative trip reset point. Accurate IR channel setpoints are difficult to obtain when the initial channel adjustments are made at low power levels (less than 30%). The initial settings for the IR channels were taken from the precautions, limitations, and setpoints (PLS)
document provided by Westinghouse (W), the Nuclear Steam Supply System (NSSS) vendor.
Because of the inaccuracies associated with the initial adjustments, additional conservative values for use in determining the setpoints were provided to the licensee in W 1etter SH-0894 dated January 12, 1987.
The IR high neutron flux trip setpoint is set for a current equivalent to 25% power.
The trip was blocked by permissive P-10 (power range low power permissive) which allows manual block of the IR trip and other low power trips when power reaches approximately 10% increasing.
During power reductions the IR trip automatically resets at approximately 12.5% power.
Permissive P-10 automatically resets and reinstates the low power trips when power decreases below 10%. Because both IR channels were not accurately set during the initial adjustments their trip resets occurred at less than 10% power instead of 12.5% power. Permissive P-10 reset before one of the IR channels (NI-36) reset.
The high neutron flux signal was received on NI-36 and thereby caused the trip. This item will be discussed in more detail when the licensee submits their licensee event report.
11.
IE Bulletin 85-01 and Other Followup (92703)
The inspector reviewed the licensee's response dated April 29, 1986, to IE Bulletin 85-01, Steam Binding of Auxiliary Feedwater Pumps. The inspector reviewed the following items discussed in the response.
Monitoring of fluid conditions within the auxiliary feedwater (AFW)
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system is performed through the use of strap-on thermocouples which are installed on the AFW system discharge lines to monitor system tempera-t ture. Should the AFW discharge line temperature exceed the established setpoint the condition will be alarmed in the main control room.
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! Abnormal L _ operating procedure (AOP) 010,, Feedwater Malfunctions,- which :
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addresses--symptoms of AFW system steam binding and corrective actions to restore the affected components.to operational status. A0P-010 has-been approved for use by the licensee.
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. Training 1 received by operations personnel) on ' recognition and recovery
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actions for AFW system steam binding was also. reviewed. These items-twere ' addressed in a lesson plan which covered AFW pump steam binding
'and AOP-010.- The lesson plan was completed September 30. 1986.
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Independent.of the above effort, NRC personnel also' inspected the AFW system and performed.'a walk through of A0P-010 with licensee operations personnel.
Questions resulting -from this effort are documented in NRC ~ Inspection-Reports 400/86-76-and 400/87-04.
Based on the above inspection efforts, the inspector determined that the licensee's response of April 29, 1986, to IE Bulletin 85-01 adequately addressed the requirements of the bulletin. Measures established to monitor AFW system temperatures, procedures which address recognizing and recovering
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. from steam binding should it occur, and material provided for operator =
ctraining~all~ appear to be adequate.
No violations or deviations were identified in the areas inspected.
(Closed) Inspector Followup Item 400/85-16-07, Special Low Power Testing to Satisfy TMI Action Item 1.G. I.
TMI Action Item 1.G 1, Training During-Low-Power Testing, is addressed in the Safety Evaluation Report (SER) dated November ~1983. Testing was added to meet the requirements of NUREG-0694, Item 1.G.1 Natural Circulation.
Natural Circulation Testing was
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performed _in accordance with the procedure 1-9103-S-23, (Revision 1). Data obtained during stable natural circulation operation were compared with -
specified conditions to those values obtained at a proto-type design plant (North Anna) for which equivalent tests.have been conducted.
Reactor coolant-system loop AT reading for natural circulation flow was verified to be within the acceptable range between 32*F - 47'F.
If the AT acceptance criterion had not been satisfied, additional operator training would have been required to close TM1 Action Item 1.G.1 during normal requalification periods. Based on the RCS loop AT reading this item is closed.
Attachments s
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ATTACHMENT 1
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PARAMETER LIST
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Unit Identification:
Plant Name SHEARON HARRIS Unit Number
Docket Number 50-400 Nuclear Steam System Supplier Westinghoase Vessel and Piping:
Volume 7982.3 cubic feet Pressurizer:
Level Units
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Temperature Compensated No Calibration Curve Slope 464.76 pounds per %
Upper Level Limi 100 %
Lower level Limit 0%
Relief Relief Tank Volume Control Tank:
Level Units
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Calibration Curve Slope 116.52 pounds per %
Upper Level Limit 100 %
Lower level limit O%
Geometric Method Available No Drain Tank:
Level Units
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Calibration Curve Slope 29.11 pounds per %
Upper Level Limit 80 %
Lower level limit 30 %
Geometric Method Available No Relief Tanks Level Units
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Calibration Curve Slope 887.55 pounds per %
Upper Level Limit
'80 %
Lower level limit 30 %
Geometric Method Available No
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ATTACHMENT 2
NRC
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INDEPENDENT MEASUREMENTS PROGRAM REACTOR COOLING SYSTEM LEAK RATES STATION: SHEARON HARRIS TEST DATE : TEST 1 UNIT
- 1 START TIME: 0244 DOCKET : 50-400 DURATION 1.317 hours0.00367 days <br />0.0881 hours <br />5.241402e-4 weeks <br />1.206185e-4 months <br /> TEST DATA
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Initial Final System Parameters Pressure, psia 2250 2250 T Ave, degrees F 558 557.3 Water Levels Pressurizer, %
24.3
Relief Tank, %
71 Volume Control Tank, %
56.8 55.1 Drain Tank, %
41 Water Charged = 0 gal Water Drained = 0 gal TEST RESULTS Change in Water Inventory in pounds:
Vessel'& Piping 373 Relief Tank (1)
O Pressurizer-604 Drain Tank (1)
O Volume Control Tank (1),-198
Less: Water Charged
Collected Leakage O
Plus: Water Drained
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Cooling System-429 Leak Rates in gpm (3):
Gross 0.65 Identified 0.00 Unidentified 0.65 (1)
Determined from tank calibration curve.
(2)
Determined from tank dimensions.
(3)
The density used for converting inventory change to leak rate was 62.31 pounds / cubic foot based on standard conditions.
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