IR 05000443/1987005
| ML20204E833 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/20/1987 |
| From: | Eselgroth P, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20204E806 | List: |
| References | |
| 50-443-87-05, 50-443-87-5, NUDOCS 8703260081 | |
| Download: ML20204E833 (7) | |
Text
-
.
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-443/87-05 Docket No. 50-443 License No. NPF - 56 Licensee:
Public Service of New Hampshire Facility Name:
Seabrook Station, Unit 1 Inspection At: Seabrook, New Hampshire Inspection Conducted:
February 17-24, 1987 Inspectors:
Td C. hm 3 3/O Peter Wen, Reactor Engineer date f
3/F/87 Approved by:
Peter Ese'1gr
, Chief, TPS, OB,DRS date Inspection Summary:
Inspection on February 17-24, 1987 (Inspection Report Number 50-443/87-05)
Areas Inspected: Startup Test Program review, post-core hot functional test witnessing and test result review.
Results: No violations were identified.
Note:
For acronyms, not identified, refer to NUREG-0544, " Handbook of Acronyms and Initialisms."
e703260001 870320 PDR ADOCK 0500
0
.
.
Details
_
1.0 Persons Contacted New Hampshire Yankee (NHY)
- J. Cooney, Technical Projects Manager
- P. Gurney, Reactor Engineering Department Supervisor R. Gwinn, Shift Test Director
- P. Hannes, System Supervisor
- G. Kann, Program Support Manager
- G. Kline, System Support Manager
- L. Rau, Reactor Startup Supervisor
- W. Temple, Licensing Coordinator T. Waechter, Shift Test Director T. Webster, System Engineer United States Nuclear Regulatory Commission (U.S.NRC)
A. Cerne, Senior Resident Inspector
- D. Ruscitto,, Resident Inspector The inspectors also contacted other administrative and technical licensee personnel during the course of the inpsection.
Denotes those present at the February 24, 1987 Exit Meeting
2.0 Post-Core Hot Functional Testing 2.1 Startup Test Program Or: February 10, 1987, the licensee began a heat up of the plant using Reactor Coolant Pump (RCP) heat. The plant achieved normal operating
temperature and pressure conditions (557 F, 2235 psig) on February 16, 1987. The licensee conducted the Post-Core Hot Functional Tests at this test plateau in accordance with the recommended test sequence of
procedure 1-ST-1, Startup Program Administration. At the conclusion of the inspection on February 24, 1987, the licensee had completed all primary system Hot Functional Tests with the exception of the Reactor Coolant System (RCS) Flow Coastdown Test (1-ST-12). Because of the test results of the Emergency Feedwater (EF) Pump Turbine Quick Start Test (1-ST-53), (see Section 2.2) many secondary system tests have not yet been completed.
These include the following tests:
EF Pump Turbine Stop Valve Trip Dynamic Response Test (1-ST-51), Turbine Driven EF Start Verification Test (1-ST-53), Steam Generator Blowdown (1-ST-54), Steam Dump System Test (1-ST-55), and Steady State Vibration Test (1-ST-56). The NRC inspection coverage for the remaining Post-Core Hot Functional Tests will be addressed in other inspection reports.
.. -
-
-
.
-
.
.
.
2.2 Test Witnessing
-
At various times during the inspection period, the inspector wit-nessed 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:
.
--l-ST-5, Control Rod Drive Mechanism Operational Test
--l-St-6, Rod Centrol System Test
--l-ST-7, Rod Drop Time Measurement
--l-ST 3, Rod Position Indication
--l-ST-9, Pressurizer Spray and Heater Capability
--l-ST-10, RTD Bypass Loop Flow Verification
--l-ST-11, RCS Flow Measurement
--l-ST-t4.2, RTD Cross Calibration
--l-ST-33, Turbine Driven EF Start Verification Test Tests were observed for the following areas:
--
Tests were conducted in accordance with the approved test progedures; 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 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 test result i
preliminary evaluations are described below.
Control Rod Drive Mechanism (CRDM) Operational Test (1-ST-5)
This test was completed on February 18, 1987.
Test results indicated that each CRDM's coil timing sequence and measured trace shape were
normal. The completion of this test demonstrated the operability of CRDMs.
No test exceptions or field changes were written against the test procedure.
However, some noise pulses were picked up by the i
CRDM Acoustic Monitoring System.
The licensee and his reactor vendor
(Westinghouse) are investigating the cause of these noise pulses.
The NRC inspector will follow this in a future inspection.
_ _ _ _
._
_ _ _ _
_
__
,_
_
__ -
-
__.
.
Digital Rod Position Indication (DRPI) Test (1-ST-8)
This test was performed on February 19, 1987.
Each bank of shutdown and control rods was in.dividually withdrawn in pre-specified increments to 228 steps. At each increment, the DRPI on the main control board was compared to the group step counter and plant computer output.
Test results indicated that the DRPI system provided accurate indications of rod position and agreed with the group steo counters within the acceptance criterion of 4 steps. The control bank group step counters also agreed with the rod control pulse-to-analog converter within the acceptance criterion of one step. However, a software problem in the plant computer was identified during this test. This software problem resulted in the demand indications for Control. Bank A through Control Bank D not being properly updated.
The startup test group has issued a work request to correct this problem. The NRC inspector will follow this in a future inspection.
Rod Drop Time Test (1-ST-7)
All 57 control rods were dropped at hot, full flow conditions. The rod drop times for each rod satisfied the acceptance criterion of less than 2.2 seconds, with the average drop time of 1.39 seconds.
Two control rods (M-12 and L-5) with both drop times of 1.36 seconds fell outside the two sigma (standard deviation) limit. These two rods were re-dropped three additional times each. The result from the re-drop test showed consistent and repeatable results.
All rod drop traces showed no evidence of binding. However, traces from rods B-10 and P-6 exhibited unusual responses. The licensee reactor angineering group is investigating the cause. The NRC in-spector will follow this in a future inspection.
Rod Control System Test (1-ST-6)
The Rod Control System was tested on February 23, 1987, to verify satis-factory performance of the required control and indication functions. The bank overlap operation was also verified during this test. All test acceptance criteria were met.
Pressurizer Spray and Heater Capability (1-ST-9)
The purpose of this test was to determine the rate of pressure reduction caused by the opening of both pressucizer spray valves (PCV-455A, PCV-4558)
and the rate of pressure increase caused by the operation of all the pressurizer heaters.
This test was performed on February 21, 1987, the preliminary test results indicated that the pressurizer responses were within the allowable bands as recommended by the reactor vendor (Westing-house).
,
L
'
__ _
_
_ _ _ _ _. _ _ _ -
- _ __
_ _ _
__
__,
__
_ _ _ _
_ -. _. _ _
_
.-
.
.
Additional information on the single spray loop operation was obtained on February 22, 1987. The data is intended for the operator training enhance-ment.
On February 18, 1987, t'he licensee performed the prsssurizer power oper-ated relief valve response time test. The measured response times were:
Valve Measured Time Allowable Time PCV-456A Opening: 0.27 Sec.
0.95 Sec.
Closing: 0.83 Sec 3.0 Sec.
PCV-456B Opening: 0.3 Sec.
0.95 Sec.
Closing:
0.83 Sec.
3.0 Sec.
These response times were well within the test acceptance criteria.
This test result satisfactorily addressed the license condition stipulated in Attachment 1 to NPF-56, Item 2.a.(3).
RTD Bypass Loop Flow Verification (1-ST-10)
This test was performed on February 19, 1987. The purpose of this test is to verify that both hot and cold RTD bypass loop flow trans-port times in each RCS loop are within 1.0 second.
The results indicated that the flows through cold leg RTD bypass loop were consistently too small for all four RCS loops as shown in the following comparisons:
Measured Minimum *
Loop No.
Leg Flow (gpm)
Required Flow (gpm)
Hot 154 76.23 Cold
47.7.3
Hot 157 77.06 Cold
48.29
Hot 151 77.88 Cold
48.0
Hot 162 76.81 Cold
51.49
- Based on actual pipe volumes and one second transport time.
The licensee is planning to correct this problem by increasing the orifice size in the cold RTD bypass loop. Upon completion of this modification, a new 1-ST-10 Test will be performed.
The NRC inspector will follow this in a future inspectio _ -.
.
.
.
6.
Reactor Coolant System Flow Measurement (1-ST-11)
This test was performed at hot, no-load conditions on February 17, 1987. The measure.d flows from the RCS elbow flow taps are as follows:
RCS Loop Average Loop Flow (gpm)
.
Loop 1 109,395 Loop 2 109,416 Loop 3 105,415 Loop 4 109,766 Acceptance Criterion Total 433,992
>391,000 The measured total RCS flow met the test acceptance criterion.
Turbine Driven EF Start Verification Test (1-ST-53)
The licensee previously performed the EF System test during the Preoperation Hot Functional Test in October, 1985. During the initial cold start test, water hammer occurred which damaged
,
several pipe supports. The EF Pump Turbine steam supply piping, EF Pump recirculation piping and system control have since been modified. The required tests were rescheduled to be performed during this Post-Core Hot Function Testing period.
.
The licensee first performed the 1-ST-53, Section 6.1 test (Steam Supply from "A" Steam Generator) on February 19, 1987. Three NRC inspectors observed the test in the control room, main steam pipe chase and EF pump house, respectively. Although the test failed because the turbine speed did not reach the rated speed within the test acceptance criterion of 60 seconds, the re-designed EF system appeared to be functioning well.
No water hammer events were ob-served. The inspector also performed an EF System walkdown at the completion of the test, and observed no abnormality in the piping system.
Various sections of 1-ST-53 were performed over the period from February 19 to 22, 1987. Problems identified during these tests were immediately corrected.
On February 22, 1987, the licensee performed the Section 6.3 test
-
(Steam Supply from both "A and "B" Steam Generators). The test re-suits indicated that the steam inlet pressure to the turbine was low.
The test was subsequently suspended. At the conclusion of this in-spection, the licensee test engineer is still investigating the cause of'the low steam inlet pressure in the EF Pump Turbine. The NRC in-spector will follow up this problem.
-
-
-
-
..
.
2.3 Summary All test results as verified by inspector direct observations indi-cated that overall test acceptance criteria have been met or proper test deficiencies were documented and followed up by the licensee.
3.0 Independent Calculations / Verifications The inspector performed independent calculations, and verified that the licensee's RCS flow measurement (1-ST-11) and RTD Bypass flow calculation (1-ST-10) were correct.
4.0 QA/QC Interface Throughout the entire inspection period, the inspector noted that QA/QC provided 24-hour shift test coverage. More than 25 QC surveillance reports have been generated since the plant heat up on February 10, 1987.
Sampling review of these surveillance reports indicated that QA/QC's coverage on the startup test program was thorough and comprehensive.
5.0 Exit Meeting An exit meeting was held on February 24, 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.
Based on the NRC Region I review of this report and discussions held with the licensee representatives at the exit, it was determined that this report does not contain information subject to 10 CFR 2.790 restric-tions.