IR 05000338/1980004
| ML19309H368 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/24/1980 |
| From: | Kellogg P, Webster E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19309H366 | List: |
| References | |
| 50-338-80-04, 50-338-80-4, NUDOCS 8005130106 | |
| Download: ML19309H368 (4) | |
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d UNITED STATES
NUCLEAR REGULATORY COMMeSSION
g a REGION 11 Os [[ o 101 MARIETTA ST., N.W., SUITE.s 100 ATLANTA, GEORGIA 30303 . Report Nos. 50-338/80-04 Licensee: Virginia Electric and Power Company P. O. Box 26666 , ) Richmond, VA Facility Name: North Anna Docket No. 50-338 i License No. NPF-4 i ) Inspection at North A na 'te near Mineral, Virginia
Inspected by: - 2,, v
If E. H. /Webgt,4r/M { Date Sign d Approved by: g "/ / N/ P. J eglfggi AcWSection Chief, RONS Branch Ddte Signed SUMMARY ' Inspection on January 22-25, 1980 Areas Inspected This routine, unannounced inspection involved 18 inspector-hours on site in the areas of licensee event reports, previous inspector concerns, and license conditions.
It also included 5 inspector hours at VEPCO corporate office in Richmond, Virginia in the area of a recent reportable event.
Results Of the 4 areas inspected, no new items of noncompliance or deviations were identified.
k 4 _ - - . DETAILS 1.
Persons Contacted Licensee Emplofees W. R. Cartwright, Station Manager
- J. D. Kellams, Superintendent Operations
- E. R. Smith, Superintendnet Technical Services
- E. W. Harrell, Superintendent Maintenance
- J. W. Ogren, Supervisor Administrative Services
- D. L. Smith, Resident QC Engineer H. W. Burress, Jr., Assistant Engineer E. I. Lozito, Director of Nuclear Fuel Operations M. L. Smith, Nuclear Fuel Engineering M. L. Bowling, Director Nuclear Fuels Engineering Other licensee employees contacted included 9 corporate engineers, 2 techni-cians, 3 operators, and 3 office personnel.
Other Organizations Stone and Webster Engineering Corporation (S&W) R. J. Daly, Lead Advisory Engineer NRC Resident Inspectors
M. S. Kidd A. P. Tattersall
- Attended exit interview.
2.
Exit Interview The inspection scope and findings were summarized on January 25, 1980, with those persons indicated in Paragraph 1 above.
! 3.
Licensee Action on Previous Inspection Findings i Not inspected.
' 4.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve noncompliance or deviations. New unresolved items identified during this inspection are ' discussed in paragraph 8.d.
! 5.
Plant Status i l l At the time of this inspection, Unit I was operating at 30% nominal power, ' [ conducting steam generator chemical soak, prior to escalating power following I I ! . - -
.... -...,..... .. - - -. ~ - - - - - - - - -. - . . - l e-2-a refueling outage. Due to a quadrant power tilt problem, discussed further in paragraph 7, reactor physics data was being collected at various control rod positions.
6.
Licensee Event Report Review The following LERs were reviewed to verify that reporting requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, and the LER forms were complete.
Addi*,ionally, for those reports identified by asterisk, a more detailed review was performed to verify that the licensee had reviewed the events, corrective actions had been taken, no unreviewed safety questions were involved, and violations of regulations or licensee / technical specification conditions had been identified.
a.
LER 79-135, dated October 29, 1979: Timers for three component actuators found incorrectly set.
This report was never received in the Region II office, however, a copy received by the NRC resident inspector was used for this review. The inspector further verified that LER 79-135 was received by the NRC Office of Management and Program Analysis with the same data.
LER 79-135 was signed and forwarded by the licensee on October 29, ' 1979, reporting an event which occurred September 25, 1979. This report was forwarded pursuant to Technical Specification 6.9.1.9 which requires reports of the types of events specified be submitted within 30 days of the occurrence of the event to the Director of the NRC Regional Office. Since this report was submitted 34 days following the event, this report is in noncompliance with Technical Specification 6.9.1.9 and is similar to an item of noncompliance identified in Inspection Report 50-338/79-50.
Whereas this item is identified prior to the licensee's response to Report 79-50, this item is reported as another example of the noncompliance (Infraction 338/79-50-07) identified in Inspection Report 50-338/79-50.
b.
LER 79-145/03L-0, dated December 4, 1979: The residual heat removal system was inoperable in Mode 6 during testing. Licensee identified followup action, to conduct a design change to the solid state pro +ec-tion systems circuitry shall be followed up in future inspections (338/80-04-01).
c.
LER 79-146/03L-0, dated December 12, 1979: Pressurizer level channel III found to be out of calibration.
d.
LER 79-154/03L-0, dated December 27, 1979: "B" steam generator support temperature indicator found to be out of calibration.
e.
LER 79-155/031-0, cated December 27, 1979: Pressurizer safety valves relief setpoints found to be too low during testing.
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-3- , 7.
Quadrant Power Tilt - On January 19, the licensee informed the NRC of an 8 percent qu.sdrant power tilt noted at 3 percent power, thermal. The licensee met with URC on January 24 in Richmond, Virginia to discuss the cause and corrective actions , to be taken by the licensee.
M. Dunnenfeld, R. Alexion, and M. Chatterton from NRR attended with A. Tattersall and this inspector from IE.
The licensee discussed a recently completed computer model based on an uneven core burnout from Cycle I which showed a 6*F higher outlet temperature in the southwest quadrant of the core located at the "A" reactor coolant loop nozzie. Other data confirmed a 4 percent to 7 percent lower flow rate in loop "A" exists which exceeds licensee requirementa for flow, but results in higher burnup rate in the southwest quadrant of the core. The resultant quadrant tilt calculated for cycle 2 closely approximated the tilt actually seen at the plant at low power. The predicted quadrant power tilt, based ' on this model decreased to less than 2 percent at 50 percent power and to less than 1 percent at 100 percent power.
During plant testing at low power, the licensee determined that the F H limits of Technical Specification 3.2.3 are exceeded at low power when the plant is operated at the control rod insertion limits of Technical Specifi-cation 3.1.3.6.
As a result, the licensee has established more conservative rod insertion limits as an administrative control. The revised limits were reviewed by the inspector and appeared to be conservative based on data taken at 3 percent and 30 percent power with control rods at the revised values.
The licensee committed to take the following action, which subsequent to this inspection have been verified complete.
Complete incore flux mapping at 30 percent power and compare this data a.
to predicted values from the computer model.
b.
Compute F at 30 percent power with the control rods at the revised H control rod insertion limits to assure compliance with Technical . Specification 3.2.3.
$ Report the results of the above listed tests to NRC and receive NRC c.
' concurrence prior to raising reactor power above 30 percent nominal power.
d.
Conduct flux mapping at 50 percent and 100 percent power, comparing results to the computer predicted values aad report these results to NRC.
The licensee also committed to utilize the revised administrative limits in . lieu of the technical specification requirements of Technical Specification 3.1.3.6 concerning rod insertion limits until a reactor physics analysis is compiled and a revision to technical specifications, as requested by the , licensee, is made. This item shall be followed up in future inspections (338/80-04-02). Further.IE review of reactor physics test results and its correlation to computer predictions shall be followed up in review of LER , 80-17/01T-0.
t a y -:
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Previous Inspector Concerns and Licensee Commitments (Closed) Open item (338/79-48-04) - Fuel rod burst calculations (LER a.
79-143). Amendment 16 to license NPF-4 dated December 28, 1979, included a safety evaluation of the increased clad heatup rates analyzed by Westinghouse in a revised LOCA-ECCS model. As a result of the , increase in peak centerline temperatures analyzed due to rod burst, amendment 16 changed Technical Specification 3.2.2 limits on FQ Form , , 2.21 to 2.10.
Licensee management was informed that based on this ! change and the SER of it in Amendment 16, this item is closed.
b.
(Closed) Open item (79-48-05) - Negative rate trip analysis error (LER 79-152) and open item (79-48-06) Reactivity profile errors during dropped rod events (LER 79-153). Based on correspondence between Westinghouse and NRC in November 1979 which discussed solution to these analysis errors, the inspector found the requirements of Operating Procedure (0P) 2.1, Unit Power Operation, Step 4.24, to satisfactorily mitigate these analysis errors. This step requires D bank rods to be in manual control and above the rod insertion limits , above 90 percent power, unless D bank is above 215 steps, in which case they may be in automatic control. This action is prescribed by . Westinghouse in their letter NS-TMA-2162 of November 15, 1979, and has ' been satisfactorily reviewed by NRC.
(Closed) Open item (79-30-01) - Inconsistent technical specification c.
requirements for charcoal filter testing. Technical Specifications 4.6.4.3, 4.7.7.1, and 4.7.8.1 were all modified by Amendment 16 license NPF-4 dated December 28, 1979, to allow 1.0 percent from leakage during tests vice the 0.5 percent leakage requirement of Regulatory Guide 1.52, Revision 2.
, d.
License condition 2D3)b, Amendment 3 to License NPF-4 dated April 1, 1978 - installation of a long-term means of protection against reactor coolant system over-pressurization. The inspector reviewed the documen-tation concerning design change 78-44 and discussed it with licensee staff and Stone and Webster personnel. At the time of the inspection, shortly after restart of Unit I from its first refueling outage, the .
installation of the overpressure protection system was documented as complete. However, the final QC checklist indicated that two loop , checks and the final functional testing were not complete on the , system. Further discussions indicated that instrument calibration of one of the nitrogen pressure transmitters was all that was required to complete the package, and that no overall system functional test had been conducted or was planned. Operability of the RCS overpressure protection system was not verified by the inspector and'shall be followed up in future inspections. Since the license condition required this system prior to restart from the first refueling outage, this is considered unresolved (338/80-04-03).
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