IR 05000413/1987057
| ML20235W662 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Catawba |
| Issue date: | 10/15/1987 |
| From: | Eapen P, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235W655 | List: |
| References | |
| 50-412-87-57, NUDOCS 8710160348 | |
| Download: ML20235W662 (6) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-412/87-57 Docket No.
50-412 License No. NPF-54 Licensee:.Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 Facility Name:
Beaver Valley Unit No. 2 l
Inspection At:
Shippingport, Pennsylvania Inspection Conducted:
August 17-27, 1987 I
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Inspectors:
P. C. Wen, Reactor Engineer, DRS date
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Approved by:
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/ 0/5-/[2 Dr. P. K. Eapen, C ief date'
Special Test Programs Section, EB, DRS Inspection Summary:
Inspection on Augast.17-27, 1987 (Inspection Report l
Number 50-412/87-57)
Areas Inspected:
Startup Test Program review, Power Ascension Test witnessing and test-result review.
Results: No violations were identified.
Note:
For acronyms not defined refer to NUREG 0544, " Handbook of Acronyms and Initialisms."
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I DETAILS 1.0 Persons Contacted Duquesne Light Company
- J. Godleski, Senior Test Engineer R. W. Huston, Reactor Engineer
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- J. Johns, Supervisor, QA Surveillance
- E. L. Martin, Compliance Engineer
- T. F. McGourty, Principle Engineer
- F. D. Schuster, Operations-Supervisor
- R. G. Williams, Principle Engineer
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- D. Szves, Compliance Engineer
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- T. G. Zyra, Director, Site Test and Plant Performance U.S. Nuclear Regulatory Commission
- J.'Beall,SeniorRe$1dentInspector L. Prividy,. Resident Inspector
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The inspector also contacted other administrative and technical licensee personnel during the course of the inspection.
- Denotes those present at the exit meeting held on August 27, 1987.
2.0 Power Ascension Testing 2.1 Startup Test Program The NRC Lifted the 5's power restriction on August 14, 1987. After completing the Mode 1 (reactor power greater than 5%) Startup Testing Checklist, the licensee plant management authorized the unit to proceed for Power Ascension Testing.
Entry into Mode 1 was accomplished on August 15, 1987. The generator was placed on the
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grid on August 17, 1987.
During this inspection period, the
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iicensee was conducting Power Ascension Testing up to the 30% power plateau.
l Two unplanned reactor trips occurred during this inspection period.
These events are discussed below:
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a.
At 0543 hours0.00628 days <br />0.151 hours <br />8.978175e-4 weeks <br />2.066115e-4 months <br /> on August 18, 1987, the generator tripped from 20%
power due to a power range negative neutron flux rate.
Prior to the reactor trip, an I&C technician was working in power cabinet
1 AC, replacing the overvoltage protector on the No. I power i
supply. While the work was in progress, he inadvertently shorted I
the No. 2 power supply.
The loss of both power supplies caused
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the control rods to drop.
Plant systems responded as designed.
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b.
On August 25, 1987, a spurious signal from the turbine electronic overspeed trip circuitry caused the turbine to trip at 1357 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.163385e-4 months <br /> while the reactor was at 30% power.
No immediate reactor trip
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occurred since at the time the reactor power was below permissive P-9 (~ 49% power).
Prior to this event, the 'C'
steam generator-already had a steam flow /feedwater flow mismatch bistable tripped
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due to an inoperable feedwater flow transmitter (2 FWS*FT*496).
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The turbine trip caused steam generator's water level to shrink.
Six seconds later, the low steam generator water level in coincidence with the steam flow /feedwater flow mismatch resulted in a reactor trip.
Immediately following the turbine / reactor trip, the unit's i
three reactor coolant pumps (RCPs) tripped due to bus under frequency caused by faulty relays.
These relays did not function properly during fast bus transfer to offsite power following j
loss of unit power supply. When the three RCPs were lost, the unit went into natural circulation as designed. Approximately S0 minutes later, the RCPs were restarted.
The inspector independently verified the computer printed plant parameters, and noted the cooldown rate to be about 38 F/hr'which
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was well below 100 F/hr, the TS limit.
In this event, the Reactor protection system functioned as designed.
The licensee took corrective action by replacing the faulty relays.
Fast transfer to offsite power was tested with acceptable results.
In general, the licensee conducted this phase of the power ascension test in accordance with IST-2.01A.10 "Startup Testing Program." This startup test administrative procedure provided a logic and comprehensive test sequence.
However, at the end of this inspection period i
(August 27, 1987), the licensee informed the'NRC that they would like to deviate from the originally established test program. A temporary deferral of such major transient tests as the MSIV closure test and the loss of offsite power test at 30% power level was proposed.
These tests were planned to be performed after the completion of the 50% power non-transient type tests.
This test program change was reviewed by the Joint Test Group, On-Site Safety Committee, and approved by the plant manager. The NRR project and technical reviewers were fully informed.
The subject was discussed and concurred upon by the NRR reviewers.
2. 2 Test Witnessing At various times during the inspection period, the inspector witnessed testing in progress on a sampling basis and evaluated the completed power ascension test.
The tests witnessed and test results evaluated included:
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IST-2.01A.07, Automatic Reactor Control Test
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IST-2.02.06, Thermal Power Calv.imetric'
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50V-2.268.02, Turbine Overspeed Trip Test
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IST-2.04.02, Shutdown from Outside the Control Room Test
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0ST-2.6.2, Reactor Coolant System Leak Rate Test
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Tests were observed for the attributes listed in Inspection Report 50-412/87-55, Section 2.2.
Details relating to those tests wi_tnessed__._
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and preliminary test result evaluations are described below,.
a.
Automatic Reactor Control Test (IST-2.01A.07)
The purpose of this test was to' verify the performance of the automatic reactor control system in maintaining reactor coolant average temperature, Tavg, within acceptable steady-state limits (+or-1.5 F of Tref).
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The test was performed on August 25, 1987 with the reactor power at 30% level. The test results indicated that the plant responded as expected.
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Thermal Power Calorimetric (IST-2.02.06)
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The purpose of this test was to determine, at selected pcuer levels, plant thermal power by means of manual calorimetric calculations.
These calculated values were used as input to the adjustment of the power range instrumentation.
The inspector reviewed results for tests performed on August 22, 1987, and subsequently verified through a control room tour, that power range readings agreed with calculated core thermal power. No discrepancies were identified.
The inspector also reviewed the technical content of the licensee's calculation method. The inspector noted that the feedwater enthalpy was conservatively taken as that for saturated liquid in lieu of normal subcooled liquid.
The results from the inspector's independent calculation agreed reasonably well with the licensee's results.
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c.
Turbine Overspeed Trip Test (SOV - 2.268.02)
The purpose of this test was to demonstrate the capability of the turbine generator to consistently trip at acceptable speeds during an overspeed condition.
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4 This test was originally performed on August 20, 1987, with unsatisfactory results. After. adjusting the mechanical overspeed weight, this test was reperformed on August 22, 1987. Test results indicated that mechanical overspeed trip and electrical overspeed controller functions
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However, the electrical overspeed pickup was found
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to be defective and could not be changed because of inaccessibility at the existing plant condition.
The retest of the electrical overspeed trip function-will be performed after.the mid-September mini outage.
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Shutdown From Outside The Control Room Test (IST-2.04.02)
The purpose of this test was to demonstrate that the plant can be safely shutdown from outside the control room, and maintained in a Hot Standby (Mode 3) condition from the aiternate shutdown panel (ASP).
and the Emergency Shutdown Panel.
The ability to cool the plant to Hot Shutdown (Mode 4) condition from'outside the control room was successfully demonstrated during the previous hot functional test (P0.2.04.02) and was not repeated in this test.
The test was performed on August 21, 1987.
The NRC inspectors stationed at. control room and remote shutdown panels observed the test in its entirety. Although minor panel labeling problems were identified during transfer of control from the ASP to the control room, the operations crew performed very well in this test.
They successfully demonstrated that the plant can be maintained in Hot Standby, and controlled from the ASP and from the Emergency Shutdown Panel.
Test results met test acceptance criteria.
e.
Reactor Coolant System Leak Rate Test (OST-2.6.2)
The inspector selectively reviewed RCS Leakrate data during this inspection period.
The inspector independently calculated the following RCS leakrate data using licensee input data and a methodology as described in NUREG-1107.
The calculated results are compared i
below:
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Test Date Leakrate Calculation Calculation 8/19/87 Total Leakrate 1.06 0.67
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Identified Leakrate 0.48 0.51 I
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Unidentified Leakrate 0,58 0.16 l
8/23/87 Total Leakrate 1.84 1.56
l Identified Leakrate 1.12 1.12 l
Unidentified Leakrate 0.72 0.44 l
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The difference between the inspector and the licensee's calculation was primarily due to using different methods for the Tavg change correction term. This subject was discussed with a. licensee representative.
The inspector will follow up on this in a future
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inspection, when more plant steady-state data become available.
The leakrate results from both licensee and inspector's calculation met-TS 3.4.6.2 limits.
2.3 Summary Most portions of up to 30% power plateau testing have been performed.
-These tests.were conducted in accordance with approved procedures.
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Test data were properly evaluated, and test objectises were met.
Problems identified during the. tests were properly documented and
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followed up by the licensee.
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No unacceptable conditions were identified.
l 3.0 Independent Calculation The inspector performed independent calculations and verified that the licensee's RCS Leakrate calculation was within the TS limits as discussed in paragraph 2.2.e.
'4.0 QA/QC Interface The licensee QA Surveillance Group continuously provided test coverage for the startup test program.
Through direct observation and discussion with the QA supervisor, the inspector noted that QA personnel had effective interface with other departments.
No Surveillance Deficiency Reports were I
issued during this inspection period.
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i No unacceptable conditions were identified.
5.0 Exit Meeting
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An exit meeting was held on August 27, 1987 to discuss the inspection scope and findings, as detailed in this report (see paragraph 1.0 for attendees).
At no time during this inspection was written material provided to the licensee by the NRC inspector.
Based on NRC Region I review of this report and discussions with licensee representatives at the exit meeting, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
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