IR 05000400/1985052
| ML18019A572 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/03/1986 |
| From: | Burris S, Fredrickson P, Maxwell G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18019A571 | List: |
| References | |
| 50-400-85-52, IEIN-85-097, IEIN-85-97, NUDOCS 8602100326 | |
| Download: ML18019A572 (9) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report No.:
50-400/85-52 License No ~
CPPR-158
~/slFL Date Signed Date Signed
,Licensee:
Carolina Power and Light Company
'. 0.
Box 1551 Raleigh, NC 27602 Docket No.:
'50-400 I
Facility, Name:
Harris Unit
'I Inspection.,Conducted:
December 20, 1985 January 20, 1986 I I
'.I
'nspectors:
F.
Maxwe
.
P.
Bur ris, Accompanying In pecto S. Mellen and W. E.'.Holland Approved y
P.
E.
ed s n, Se tion Chief Division of 'Reactor Projects SUMMARY Scope:
This routine, announced inspection involved 188 inspector-hours on site in the areas of Preoperational Testing, Comparison of As-Built to FSAR Description, Integrated Hot Functional Test Witnessing, and Licensee Activities.
Results:
No violations or deviations were identified.
8602100326 860204 pDR ADOCK 05000400, Q
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REPORT DETAILS Persons Contacted Licensee Employees N. J. Chiangi, Manager, QA/QC Harris Plant J
~
M. Collins, Manager, Operations G.
L. Forehand, Director, QA/QC J.
L. Harness, Assistant Plant General Manager, Operations C.
S. Hinnant, Manager, Start-up L. I. Loflin, Manager, Harris Plant Engineering Support D. A. McGaw, Superintendent QA C.
L. McKenzie, Acting Director, Operations QA/QC G.
A. Myer, General Manager, Milestone Completion R.
M. Parsons, Project General Manager, Construction Confirmation Completion M. Thompson, Jr.,
Manager, Engineering Management D.
L. Tibbitts, Director, Regulatory Compliance B.
Van Metre, Manager, Harris Plant Maintenance M.
D. Vernon, Superintendent QC E. J.
Wagner, Manager, Engineering C.
C. Wagoner, Project General Manager, Construction R. A. Watson, Vice President, Harris Nuclear Project J.
L. Willis, Plant General Manager, Operations Other licensee employees contacted included technicians, operators, and engineering personnel.
Exit Interview The inspection scope and findings were summarized on January 21, 1986, with the Plant General Manager, Operations.
Four Inspector Follow-up Items were identified as follows:
"Construction Fire Brigade Drill", Paragraph 3;
"Operational Discrepancies of the Component Cooling Water System",
Paragraph 4;
"Leaking Steam Generator Relief Valves",
Paragraph 5;
and "Pressurizer Pressure Indicators",
Paragraph 5.
No written material was provided to the licensee by the resident inspectors during this reporting period.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the resident inspectors during this inspection.
Preoperational Test Program Implementation Verification (71302)
The inspectors conducted routine tours of the facility to make an independent assessment of equipment conditions, plant conditions, security, and adherence to regulatory requirements.
The tours included a
general observation of plant areas to determine if fire hazards existed; observation of activities in progress (e.g
~
, maintenance, preoperational testing, etc.)
to determine if they were being conducted in accordance with approved pr'ocedures; and observation of activities which could damage installed
equipment or instrumentation.
The tours also included evaluation of system cleanness controls and a
review of logs maintained by test groups to identify problems that may be appropriate for additional follow-up.
On January 9,
1986, the site construction fire brigade conducted a drill.
The drill began when the main control room announced that a fire was reported in construction warehouse number 3.
The site rescue squad and the fire brigade were dispatched to the drill site; approximately 15 minutes passed before the fire brigade arrived.
On January 20, the inspectors contacted the Plant General Manager, Operations, and requested an explanation for the inordinately long response time to the drill by the Construction fire brigade.
By the end of this reporting period, the
'nspectors had not received a
sati sfactory response to their inquiry.
Therefore, this will be identified.
as an Inspector Follow-up Item,
"Construction Fire Brigade Drill" 400/85-52-01.
No violations or deviations were identified in the areas inspected.
Comparison of As-Built Plant to FSAR Description (37301)
The inspectors completed a
walkdown of the accessible portions of the component cooling water system (CCW) required for Unit 1 operation.
The CCW system was inspected for conformance with component cooling system simplified flow diagrams CAR-2165-G-819 (Rev.
12),
CPL 2165-S-1319 (Rev. 2),
and CPL 2165-S-1320 (Rev.
0).
During the inspection, the following discrepancies were noted.
The FSAR Figure 9.2.2-1 (Rev.
11) was found to be not the most current revision of the EBASCO design drawing.
Loop A suction strainer was missing from the as-built piping system.
Valve 1CC-32 was missing a bonnet nut.
Valve identification tags were missing on valves 1CC49, 1CC51, 1CC107, 1CC75, 1CC99 and 1CC113.
The tag for valve 1CC108 was misplaced.
Identification tags were reversed on valves 1CC65 and 1CC66.
Identification tags for 1CC127, 1CC43 and 1CC167 needed to be, correctly fastened.
Pressure gauge 1-PI-678B for loop B
pump discharge pressure had a
broken face glass.
Several piping connections were not capped, which could compromise system cleanness.
The local temperature gauges on the RHR heat exchangers 1A-SA and 1B-SB were broken, displaying readings which were off-scal The preceding deficiencies were discussed with the licensee and are identified as an Inspector Follow-up Item, "Operational Discrepancies of the Component Cooling Water System" 400/85-52-02.
In the areas inspected, no violations or deviations were identified.
Integrated Hot Functional Test Witnessing (70314)
The integrated hot functional preoperational testing for Harris commenced on December 22, 1985.
The major purpose of this testing evolution is to conduct testing on installed systems and components.
Heatup from ambient to hot standby and control of testing during this time frame were accomplished by test instruction 2005-P01 and Operations General Procedure GP-001.
2005-P01 and GP-002 established plateaus at 150 degrees F,
250 degrees F,
350 degrees F,
450 degrees F,
and.557 degrees F,
normal operating temperature (NOT)
and listed tests that were required to be performed at each plateau.
The inspectors witnessed operations personnel completing prerequisites in preparation for plant heatup.
Operations personnel used proper procedures in addition to the test instructions for plant heatup.
Of those test procedure steps witnessed, each step was accomplished in a professional manner.
Testing accomplished at the 250 degree F
plateau involved several activities.
data were recorded including RCS pressure, temperature, flow, reactor coolant pump data, reactor vessel flange leakoff temperature, and reactor coolant system thermal expansion measurements were taken as required by test procedure 2005-P07.
At the 350 degree F plateau testing was also conducted.
Reactor coolant system instrumentation data were again recorded; reactor coolant system thermal expansion data were taken as required by 2005-P07, and reactor coolant system leakage was verified as being within proposed Technical Specification limits.
The steam generator blow down system was put into service per operating procedure OP-127.
At the 450 degree F plateau the reactor coolant system instrumentation data were again recorded.
System thermal expansion data were taken as required by 2005-P01.
Resistance temperature detector/thermocouple cross calibration data were taken as required by 2005-P03.
The inadequate core cooling monitor data were taken as required by 2005-P01 data sheet 10.9.
The operators then commenced heatup to 557 degrees F
(NOT)
and increased
'ressure to 2235 psig, normal operating pressure (NOP).
When steam generator pressure reached approximately 1092 psig (557 degrees F), three steam generator safety relief valves started leaking.
Two of the valves were for steam generator A,
and one for steam generator C.
The operators reduced the steam pressure to less than 1000 psig by reducing reactor coolant temperature.
The affected relief valves were then gagged shut to prevent their leakage.
Each steam generator has five safety relief valves, and any one of the valves is capable of protecting the steam generator from being overpressurized at hot functional test conditions.
CP&L and Westinghouse evaluated the condition of the leaking valves and indicated that the leakage probably would be corrected when the valves are set at their prescribed lift setpoints.
The setting of the valves is expected to be accompli shed prior to the completion of hot functional testing.
This will be identified as an Inspector Follow-up Item, "Leaking Steam Generator Relief Valves" 400/85-52-03.
The licensee began conducting the pressurizer pressure and level controls test (2005-P04) while at NOT and NOP, and the the following event occurred:
during performance of section 6,
step 6. 1.2, the test engineer noted that the pressurizer pressure indications recorded on data sheet 10. 1 were not within the required tolerance(two percent of the 800 psig indication band).
The test engineer halted the test and requested maintenance to perform a
troubleshooting check of the applicable circuit located in the process instrumentation cabinets and pressure transmitters.
Resolution of the problems with these pressure indicators will be identified as an Inspector Follow-up Item, "Pressurizer Pressure Indicators" 400/85-52-04.
No violations or deviations were identified in the areas inspected.
6.
Licensee Activities (40301B)
The inspectors followed up on an IE Information Notice, 85-94,
"Potential for Loss of Minimum Flow Paths Leading to Pump Damage During a
LOCA".
The licensee has evaluated this Notice and determined that, the Harris Plant does not have any emergency core cooling systems which would cause a total system failure under a single failure criteria.
The inspectors reviewed CP&L's evaluation with safety committee personnel and informed Regional management of the results of this revie i
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