IR 05000412/1987055
| ML20237L220 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/20/1987 |
| From: | Eselgroth P, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237L213 | List: |
| References | |
| 50-412-87-55, NUDOCS 8708280009 | |
| Download: ML20237L220 (8) | |
Text
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
J Report No.
50-412/87-55 Docket No.
50-412 License No.
NPF-64 Licensee: Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 Facility Name:
Beaver Valley Unit No. 2 Inspection At:
Shippingport, PA Inspection Conducted.: July 27-31, 1987
/~cA 20 87 Inspectors:
-
-
P.
en, Reactor Engineer, DRS
'dat6 j
d-Approved by:
MJ 8[ 4/87 P. W. Eselgr
, Chisf, TPS, 08, DRS
' date Inspection Summary:
Inspection on July 27-31, 1987 (Inspection Report
'
Number 50-412/87-55)
Areas Inspected:
Startup Test Program review, post-core hot functional test l
witnessing and test result review.
Results: No violations were identified.
Note:
For acronyms not defired refer to NUREG V544, " Handbook of Acronyms and Initialisms."
8708280009 870821 PDR ADOCK 05000412 G
.
l
,
DETAILS I
1.0 Persons Contacted i
Duquesne Light Company
- J. Godleski, Senior Test Engineer R. W. Huston, Reactor Engineer
- J. Johns, Supervisor, QA Surveillance
- W. S. Lacey, Plant Manager l
i K. J. Lynch, Test Engineer j
R. J. Manko, Startup Test Coordinator I
- T. P. Noonan, Assistant Plant Manager
- D. Szves, Compliance Engineer
,
- T. G. Zyra, Director, Site Test and Plant Performance
U.S. Nuclear Regulatory Commission
- J. Beall, Senior Resident Inspector i
L. Prividy, Resident Inspector
)
The inspector also contacted other administrative and technical licensee personnel during the course of the inspection, j
- Denotes those present at the exit meeting held on July 31, 1987.
2.0 Post-Core Hot Functional Testing 2.1 Startup Test Program During this inspect %n perMd, the unit remained in Mode 3 with the reactor coolant sy'; tem (RCS) temperature maintained at about 547 F.
The licensee was continuously performing the pre planned initial startup tests and preparing for initial criticality.
I On July 30, 198 an inadvertent safety injection (SI) actuation occurred while Automatic Steam Generator Water Level Control, Section A, test (S0V-2.24C.01) and Loop C Steamline Pressure Protection Channel IV (MSP-21.09-1) test were in progress. The SI was initiated because the Steam Generator 'C' steam pressure bistables were simultaneously tripped in two different channels (Channel 3 from S0V-2.24C.01 test and Channel 4 from MSP-21.09-I test).
l Approximately 2,000 gallons of Refueling Water Storage Tank (RWST)
water were injected into the RCS.
Both diesels started, and all other systems operated as expected.
The actions of the control room operators following the SI were determined to be both proper and in l
accordance with procedural requirements.
However, because this SI
'
was caused by test personnel during the conduct of testing, the need to improve the pre-test briefing was discussed with plant management.
In the exit meeting, the licensee representative stated that test j
performance in this area will be strengthened.
The inspector will
'
follow this in a future inspection.
__
___
__
.
.
2.2 Test Witnessing At various times during the inspection period, the inspector witnessed testing in progress on a sampling basis and evaluated most portions of the Post Core Hot Functional Test.
The tests witnessed and test results evaluated included:
P0-2.03.01, "Incore Thermocouple and RTD Cross Calibration."
--
P0-2.06.10, " Pressurizer Continuous Spray Flow"
--
--
IST-2.06.02, " Pressurizer Heater and Spray Capability Test".
)
.{
IST-2.06.04, " Reactor Coolant System Flow Measurement Test".
--
--
IST-2.06.03, "RTD Bypass Loop Flow Verification".
--
IST-2.01B.03, " Rod Drop Time Measurement Test".
--
IST-2.018.04, " Rod Control System Test".
IST-2.06.05, " Reactor Coolant System Flow Coastdown Test".
--
Tests were observed for the following areas:
Tests were conducted in accordance with the approved test j
--
procedures-
!
Change to the procedures were made in accordance with the
--
administrative procedure; Prior to performing each test, briefing with the test crew and
--
operation personnel were conducted and the briefing was adequate; Test prerequisites and initial conditions were met;
--
Adequate communications were established for test performance;
--
--
Operator actions were correct; and, Summary analysis was made upon completion of each test.
--
Details relating to some of those tests witnessed and preliminary test result evaluations are described below.
Incore Thermocouple and RTD Cross Calibration (PO-2.03.01)
The purpose of this test was to provide a functional check out of the RCS resistance temperature detectors (RTD's) and the incore thermocouple (TC's) and to generate isothermal cross calibration
- _ _ _ - _ _ _ - - - - - _ _ - _ - - _ _ _
-_
--
- __ _ __ _ _ _ _
_ - -
!
.j I
data for subsequent determination of individual RTD installation l
correction factors.
The test was performed at six different
'
temperature plateaus: 250 F, 350 F, 400 F, 450 F, 500 F, and 547 F.
Due to instrument (FLUKE Ohms-Converter digital multimeter [DMM])
setup error, the measured data did not indicate that isothermal conditions were being met.
Licensee test personnel continuously i
l performed troubleshooting during the entire post-core hot functional I
test period.
On July 29, 1987, the cause was traced to a push-button I
of the FLUKE DMM which was not being pushed during the tests.
This push-button separates the instrument application in a 4-wire mode (correct configuration) from a 2-wire mode (incorrect configuration).
i The. data taken at the 547 F plateau with the FLUKE DMM in the proper j
configuration confirmed the existence of the isothermal conditions.
Based on the licensee NSSS vendor's evaluation, this set of data J
represented that the RTDs accuracy is within 1.2 F, which is the total error limit assumed in the safety analysis. However, since there was only one set of valid test data obtained during the post-
,
core hot functional test, the licensee is planning to take at least 3 l
more sets of data at appropriate plant conditions.
The inspector will follow this in a future inspection.
Pressurizer Continuous Spray Flow Test (PO-2.06.10)
I The purpose of this test was to establish an optimum pressurizer spray bypass valve position in order to maintain the spray lines in a warmed condition (to minimize thermal shock on the lines when pressurizer spray is initiated).
The test results indicated that even with both bypass spray valves (2RCS*51 and 52) full open, the continuous flow through the spray line was insufficient. As a result, the Pressurizer Spray Line Low Temperature Annunicator could not be cleared during the test; indicating that the Pressurized Spray Line was below 530 F (Low i
i
!
Temperature Alarm Setpoint).
To correct this problem, the NSSS l
'
vendor recommended use of the Pressurizer back-up heater to cause i
normal spray valves opening, and subsequent heating of the spray
!
line to a temperature greater than 530 F.
Through control room tour, the inspector noted that this practice was implemented and control room operators were aware of this problem.
For the long l
term fix, the licensee is planning to perform maintenance work on
'
the bypass valves and inspect spray line orifices in the coming outage.
The inspector will follow this in a future inspection.
Pressurizer Heater and Spray Capability Test (ISI-2.06.02)
The purpose of this test was to determine the rate of pressure reduction caused by the opening of both pressurizer spray valves and the rate of pressure increase caused by the operation of all the pressurizer heaters.
l-
.
.
Test results indicated that pressurizer pressure response to opening of both normal spray valves (-3.2 psi /sec) was within the test limits (-2.0 to -3.8 psi /sec) as recommended by the NSSS vendor.
However, the pressurizer heater pressurization rate (0.17 psi /sec)
did not meet the test acceptance criteria (0.25 to 0.44 psi /sec).
An engineering evaluation was performed by the NSSS vendor, and'
concluded that there was no safety concern.
Subsequently, the test results were accepted by the Joint Test Group (JTG).
)
Reactor Coolant systam Flow Measurement (IST-2.06.04)
This test was performed at hot, no-load conditions on July 25, 1987. The measured flows from elbow flow taps are as follows:
RCS Average Loop Flow (gpm)
Run #1 Run #2 Loop 1 92,333 91,767 Loop 2 95,793 94,492 i
Loop 3 94,403 94,413 Acceptance Criterion l
l Total 282,529 280,672
>247,320 l
,
The measured total RCS flow met the test acceptance criterion.
.
RTD Bypass Loop Flow Verification (IST-2.06.03)
l The purpose of this test was to verify that both hot and cold RTD bypass loop flow transport times in each RCS loop were within 1.0 second.
This test also checked the low flow alarm setpoint for each of the RTD bypass loop flow indicating switches.
The results indica;ed that RTD bypass loop flows from loop 1 cold q
leg, loop 2 hot irg and loop 3 cold leg were less than the minimum i
required values as shown in the following comparisons:
)
i
!
l i
.
..
.
.
.
.
...
.
.
.....
.
.
. _ - _ - - - - _ _ -
i
.
i
.
Minimum *
Measured Recuired Transport i
Loop No.
Leg Flow (gpm)
Flow (gpm)
Time (sec)
(
I
Hot 115.3 97.3 0.84 j
Cold 58.7 67.3 1.15 j
Hot 102.9 104.5 1.02 Cold 75.2 70.6 0.94
Hot 112.1 99.8 0.89 Cold 63.9 69.3 1.09
- Based on actual pipe volumes and one second transport time.
)
l The inspector independently verified the above minimum required flow i
calculation, and found the licensee's calculation to be accurate.
l l
The largest transport time is in the loop 1 cold leg RTD bypass of l
1.15 seconds. This exceeds the test acceptance of 1.0 seconds by l
0.15 sec.
l An engineering evaluation was performed by the NSSS vendor, and l
recommended to accept the test result as-is.
The evaluation i
indicated that transport times greater than one second was
!
!
acceptable as long as the overall response of the overtemperature and overpower delta T reactor trip was less than tne FSAR i
requirement of 6 seconds and the respon.;e time of the associated
-
reactor trip instrumentation was less r.han the TS requirement of four seconds.
In the previous response time test of the overtemperature and overpower circuitry (analog / logic and trip breaker portions),
the actual response time was about 0.25 seconds (Loop 1) whereas the test required response time for this circuitry was 1.15 seconds.
This margin (1.15 seconds - 0.25 seconds = 0.9 seconds) is sufficient to compensate for the 0.15 seconds that the loop bypass
,
l transport delay time exceeded the acceptance criteria of 1 second.
The overall acceptability of the response times associated with the overtemperature and overpower delta T protection system will be evaluated after the plant trip from 100% power test. The inspector l
will follow this in a future inspection.
Rod Drop Time Measurement Test (IST-2.018.03)
The rod drop time measurement test at hot full flow conditions were performed on July 23-25, 1987.
The inspector reviewed test results-and noted that 46 of 48 drop times were within the two-sigma limits.
The two rods (K2 and B6) that were outside the two-sigma limit were redropped 5 times each.
The ranges of redropped times were 0.016 seconds and 0.017 seconds for K2 and B6, respectively;
I
.
I
.
!
these results are within acceptance criteria of 0,02 seconds. All 48 control rod drop times met the TS limit (s 2.2 seconds from decay of the stationary gripper voltage to dashpot' entry).
Based on the test results obtained under cold no flow, cold full I
flow and hot full flow, and earlier Unit I startup test result, the licensee determined that the rod drop test under hot, full flow was more limiting than under hot, no flow conditions.
The rod drop test
.
under hot, no flow conditions was therefore deleted from the IST 2.01B.03 test.
This deletion was reviewed by the Joint Test Group, On-site Safety Committee, and approved by the plant manager.
The inspector independently reviewed the test results and concurred with the licensee's conclusion.
The inspector also consulted with the
NRR FSAR technical reviewer, and have no further questions, i
Rod Control System Test (IST-2.018.04)
!
l The purposes of this test were (1) to verify proper control rod j
l motion, direction, rod speed and position indication in individual rod banks control mode, and (2) to verify proper operation of the l
control rod bank overlap circuitry in manual rod bank control mode.
.,
All test results met test acceptance criteria.
!
Reactor Coolant System Flow Coastdown Test (IST-2.06.051 l
'
The safety analysis performed for a postulated case of complete loss of forced reacto. coolant flow was documented in the FSAR Chapter
,
5.3.2.
The analysis result indicates that the flow coastdown following a trip of all three RCPs is adequate to prevent the l
Departure from Nucleate Boiling (DNB) from occurring.
The purpose of this test was to verify the assumptions used in the safety analysis.
This test was performed to measure the rate of change of reactor coolant flow by simultaneously tripping all three RCPs. The test-was performed on July 27, 1987.
The inspector noted that the calculated flow coastdown constant from the measured data was 14.93 seconds.
This time constant met the test acceptance criteria of 2: 11.96 seconds.
2.3 Summary Most portions of the post-core hot functional test as reviewed by the inspector through direct test observation and/or test result review indicated that tests were generally conducted in accordance with approved procedures, test data were properly evaluated, and test objectives were met.
Problems identified during the tests were I
properly documented and followed up by the licensee.
- _ _ - - _ _ _ _ _ _ - - _ _
. _ _ _ _ _ _ _ _
.
,
.
'
l
I
'
,
l There appeared to be some room for improvement in i.iic test l
personnel's pre-test briefing.
Through direct test observation, the inspector noted that some test briefings were not conducted thoroughly.
This subject was discussed with plant manager. A licensee representative stated that this area would.be strengthened in the future startup testing activities.
Licensee management was responsive to inspector observations.
Problem areas were quickly corrected and actions were taken to preclude their recurrence. Overall licensee performance in the test area can be considered effective, and are acceptable.
3.0 Independent Calculation / Verification The inspector performed independent calculations and verified that the licensee's RTD Bypass flow calculation was correct.
l 4.0 QA/QC Interface The licensee QA Surveillance Group continuously provided test coverage for the startup test program. QA activities in the startup testing area were properly documented in the QA surveillance reports.
Sampling review of these surveillance reports indicated that QA's coverage on the startup.
test program was adequate.
5.0 Exit Meeting An exit meeting was held on July 31, 1987 to discuss the inspection scope i
,
and findings, as detailed in this report (see paragraph 1.0 for I
attendees).
At no time during this inspection was written material provided to the
,
I licensee.
Based on the NRC Region I review of this report and discussions held with the licensee representatives at the exit, it was determiend that this report does not contain information subject to 10 CFR 2.790 restrictions.
!
t__
_
_
_ _ _. _. _ _ _ _ _ _
_ _ _
._ _
__.____._._____._________m