IR 05000395/1990014
| ML20043D209 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/21/1990 |
| From: | Belisle G, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20043D207 | List: |
| References | |
| 50-395-90-14, NUDOCS 9006070261 | |
| Download: ML20043D209 (5) | |
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- UNITEG) STAT ES Sa mico,D
o NUCLEAR REGULATORY COMMisslON y*
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R EGION 11
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.k 101 MARitTTA STREET.N.W.
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t ATt. ANT A. CE ORCI A 30323
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t Report No.:
50-395/90-14 Licensee: South Carolina Electric & Gas Company Columbia, SC 29218 Docket No.:
50-395 License No.: NPF-12 Facility Name:
V. C. Summer Inspection Conducted: fpril30-May4,1990
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fo Inspector:
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J. J. Lenahan " '
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DateSigned Approved by:
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h /f Ji Db/O G. A. Belisle,'ChTef
,Date' Signed Test Programs Section Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection was conducted in the areas of the containment building tendon surveillance program, testing of the main steam and pressurizer safety relief valves, and modifications to the pressurizer loop seal system and reactor building level instrumentation.
Results:
In the areas inspected, violations or deviations were not identified.
The licensee's resciution of problems identified during plant operations is identified as a strength.
Examples are modifications performed on the pressurizer safety valve loop seals, reactor building level instrumentation, and resolution of tendon surveillance problems.
Approach to resolution of problems demonstrate clear understanding of problems, and were technically sound, thorough, and timely.
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e REPORT DETAILS 1.
Persons Contacted Licensee Employees 0. Bradham, Vice-President, Nuclear Operation
- H. Donnelly, Senior Engineer, Nuclear Licensing
- A. Koon, Manager, Nuclear Licensing E. Lynch, Quality Assurance Specialist
- G. Moffatt, Manager, Maintenance
- D Moore, General Manager, Engineering Service A. Rice Licensing Engineer, Nuclear Licensing
- J. Skolds, General Manager, Nuclear Plant Operations
- G. Soult, General Manger, Operation and Maintenance J. Todd, Civil Engineer G. Williams, Mechanical Engineer J
Other licensee employees contacted during this inspection included two engineers and two technicians.
NRC Resident inspector
- L. P. Modenos, Resident inspector
- Attended exit interview 2.
ContainmentBuildingTendonSurveillance(61701)
The inspector examined quality records relating to the containment building tendon surveillance program and retensioning of the vertical tendons.
Acceptance criteria utilized by the inspector appear in
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Technical Specification (TS) 3/4.6.1.6; SCE&G procedure ' number SP-228, Surveillance of Reactor Building Post Tensioning System; and SCE&G Procurement Technical Requirements (PTR), number SC-33.
The inspector performed a detailed review of SP-228 and PTR SC-33 during an inspection conducted January 29 - February 2,1990, documented in NRC Inspection Report number 50-395/90-05.
During the current inspection.-the inspector examined the following records:
a.
Results of tensile tests performed on wire samples removed from tendon numbers D3-32, 43AC, and V-51.
All test results were-acceptable.
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b.
Results of lift off measurements for dome tendons 01-08, D1-16 through D1-19, D2-11, D2-26, 03-03, and 03-32, c.
Results of lift off measurements for horizontal tendons 7AC, 8CB, 12AC-14AC, 15BA-19BA, 32AC-34AC, 34CB, 43AC, and 46AC.
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Results of lift.off measurements performed on vertical tendons VI and V-115 prior to tendon retensioning, and records of vertical tendon retensioning, including elongation measurements and "as-left" lift off measurements.
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Records documenting regreasing of the above listed tendons, f.
Pre-surveillance stressing ram calibration records for ram numbers 9361, 9362, 9363, 9365, and 9366. The stressing rams were calibrated in accordance with Precision Surveillance Corporation Procedure, QA
12.8.G-2, Calibration - Stressing Ram / Jack.
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Results of calibrations performed on pressure gauges.
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Nonconformance report number 9N 373-001 through 9N 373-008.
All nonconformances were satisfactorily dispositioned.
Based on review of the above listed records, the inspector concluded that the tendon surveillance activities were completed in accordance with TS requirements. The containment building post tensioning system conforms to design requirements.
All testing required by the TSs and licensee's
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Quality Assurance program was completed excep)t for chemical tests of the tendon corrosion protection material (grease and the post surveillance stressing ram calibrations which were in progress during the current inspection.
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Within the areas inspected, no violations or deviations were identified.
3.
Modification and Testing of Pressurizer Safety Relief Valves (61701 and 37701)
Seat leakage through the pressurizer safety valves resulted in a loss of the water seal on the valve inlets. The loss of the water seal caused an increase in the safety valve temperature and a decrease in the set point of
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the valves, which sometimes resulted in inadvertent opening of the valves and depressurization of the reactor coolant system.
Two reactor trips occurred in 1989 due to the inadvertent opening of the valves.
The licensee implemented a design modification, MRF 21635, to drain the loop seals, and install steam seal internals in the pressurizer safety relief
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valves.
This modification should eliminate the potential for inadvertent reactor coolant system depressurization resulting from a leakage induced loss of the water loop seal.
L The inspector examined MRF 21635, including the 10 CFR 50.59 Safety Assessment, design drawings, modification prerequisites, and post modifi-cation testing requirements.
The inspector also examined MCNs A, B, and C to MRF 21635, which document minor changes to the original MRF package.
.The inspector walked down the pressurizer safety valve loop seal system and examined the new loop seal drain piping and associated pipe supports.
Field work for the modification had been completed as of the inspection l_
date except for installation of mirror insulation on the new drain piping.
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The pressurizer safety valves were modified in accordance with Drawing No. DS-C-56964-9, Nozzle Type Safety Valve, to install steam seal internals ir, the valves.
After the valves were modified, set point i
testing was performed to adjust the valves for anticipated temperature conditions during plant operation.
The valves were set to 2485 psig i 1%
per TS 3.4.2.2.
Following completion of the set point testing, a jack and lap procedure was performed on each valve to make the valves more leak tight.
In using this procedure, the valves are opened with a hydraulic jack, the valve seats are lapped slightly to remove any imperfections, and leak tested using nitrogen after the hydraulic jack is removed.
The set point of the valve is not changed during the jack and lap procedure.
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The inspector examined the set point test data and verified that the final set points met TS requirements.
The set point testing was performed at various temperatures to obtain data showing the affects of various temperatures on the valve's set points.
The final set point adjustments were made at the anticipated system temperature.
Testing was performed in accordance with SCE&G procedure STP 401.001, Pressurizer Code Safety Valve Test.
The licensee has installed resistance temperature detectors at
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various locations on the modified pressurizer safety relief valves which
will monitor the valve temperature during plant startup with the loop seals drained to ascertain that actual system temperatures are within limits used to adjust the set points for the modified valves.
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Within the areas inspected, no violations or deviations were identified.
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Main Steam Safety Relief Valve Set Poirt Testing (61701)
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The inspector examined results of testing performed on the main steam safety relief valves (SRVs) and verified compliance with TS 4.0.5 which requires that SRVs be set point tested in accordance with Section XI of
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the ASME Code during each refueling outage.
The valves are tested per procedure STP-401.002, Main Steam Line Code Safety Valves, ASME Section XI Test.
The lift setting of the valves are specified in TS Table 3.7-3.
The set points are required to be within plus or minus one percent of the
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specified lift setting.
Review of the set point test data for the surveillance performed on March 22, 1990, disclosed that 7 of the 15 SRVs were slightly outside of the tolerance.
The valves were adjusted and j
retested. The as-left set points were within the TS limits. The licensee
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documented the set point test failures on Off-Normal Occurrence Report l
90-033.
The licensee determined that this problem was not reportable.
The inspector concurs.
Corrective actions were performed immediately after the initial set point testing was completed.
The inspector also examined the results of set point testing performed on the SRVs on September 18, 1988, during the last refueling outage.
Seven of the 15 SRVs but not the same SRVs were also slightly outside the TS limits.
The as found set points fo'c the seven SRVs were 1.1 to 3.1 percent high, l
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The as left set points were within TS L
limits.
Within the areas inspected, no violations or deviations were identified.
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Modification of RHR Sump and Reactor Building Level Instrumentation
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(37701)
The inspector examined MRF 21479, which replaced the RHR sump and reactor i
building level transmitter with a common instrument which does not utilize a dry reference leg vent tube. The dry reference leg vent was subject to possible erroneous readings due to collection of condensation during l
accident conditions in the reference leg.
The new level transmitters do not require atmospheric vents.
The common level transmitter will provide data to cover two level readings:
1) the reactor buildin protection from flooding of safety-related equipment and 2)g level for the head of j
water available in the reactor building sump when the RHR pumps are taking suction from the sump.
The inspector reviewed the post modification j
testing requirements for the new level transmitters.
These requirements
are specified in Surveillance Test procedure STP 300.03, Reactor Building
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Residual Heat Removal Sump A, Level Instrument ILT 1969 Calibration and STP 300.04, Reactor Building Residual Heat Removal Sump B. Level
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Instrument ILT 1970 Calibration.
Additional routine surveillances are also performed per the requirements of TS 3/4.3.3.6.
Within the areas inspected, no violations or deviations were identified.
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Exit Interview The inspection scope and results were summarized on May 4,1990, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in dctail the inspection results.
Proprietary information is not contained in this report.
Dissenting consnents were not received from the licensee.
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