IR 05000338/1980013
| ML19331C327 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/11/1980 |
| From: | Ford E, Kellogg P, Webster E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19331C307 | List: |
| References | |
| 50-338-80-13, 50-339-80-14, NUDOCS 8008140561 | |
| Download: ML19331C327 (7) | |
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UNITED STATES
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'n NUCLEAR REGULATORY COMMISSION
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REGION II
o, 101 MARIETTA ST N.W SUITE 3100
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ATLANTA, G EoRGIA 30303 Report Nos. 50-338/80-13 and 50-339/80-14
Licensee: Virginia Elect *ic and Power Company Richmond, Virginia 23261
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Facility: North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and CPPR-78 Inspection at Nor a ite ' ear Mineral, Virginia
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Inspectors:
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E. J. Ford Date Signed
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C. Juli~a Date Sig ed Approved by:
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P. J/ Kel/l4M A W Section Ch2ef, RONS Branch W
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SUMMARY Inspection on March 24-27, 1980 Unit 1 Areas Inspected This routine unannounced inspection involved 22 inspector-hours on site in the areas of licensee event reports and previous inspection findings.
- Unit 1 Findings Of the 2 areas inspected, no items of noncompliance or deviations were identified.
Unit 2 Areas Inspected This routine unannounced inspection involved 32 inspector-hours on site in the areas of preoperational test results and previous inspection findings.
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Unit 2 Findings Of the 2 areas inspected, no items of noncompliance or deviati6ns were identified.
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DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager
- J. D. Kellams,_ Superintendent, Operations
- E. R. Smith, Jr., Superintendent, Technical Services
- E. W. Harrell, Superintendent, Maintenance
- G. A. Kann, Engineering Supervisor F. Termine11a, Engineer R. F. Anderson, Acting Operations Supervisor Other licensee employees contacted included 2 mechanics NRC Resident Inspector M. S. Kidd
- A. P. Tattersall
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on March 27, 1980, for those persons indicated in paragraph I above. The inspector concern described in paragraph 5 was discussed with licensee management at length.
It was concluded that the attention to detail required in future reviews of test procedures would be increased to preclude procedure deviations and that complete documentation of deviations would be stressed to all concerned.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Preoperational Test Results Review (Unit 2)
Completed preoperational test procedure results were reviewed for confor-mance with Nuclear Power Station Quality Assurance Manual sections 2, 5, 11, 14 and FSAR Section 14.1.1.
No items of noncompliance or deviations were noted during the review a.
of the following procedures:
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(1)
2-PO-3o.1, Revision 0, dated 6-22-18, Safety Injection System
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Functional Test Breakers and Valves
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-2-(2) 2-PO-36.3 Revision 0, 12-26-78, Safety Injection System Functional Test Pump Performance Verification (3) 2-PO-70B, Revision 0, dated 5-14-79, Emergency Component Sequencing Test During Loss of Normal Emergency Power (4)
2-PO-65.3A, Revision 0, dated 12-19-79, Turbine Trip to Reactor Trip Response Time (5)
2-PO-65.3, Revision 0, dated 7-17-78, Engineered Safety Features Equipment and Containment Isolation Valve Response Times (6)
2-PO-65.6, Revision 0, dated 6-21-78, Reactor Protection and Engineered Safety Features - Total Response Time (7)
2-PO-66.2, Revision 0, dated 5-18-79, DC Distribution System -
Station Battery Service Test The inspector noted during the review of test procedure results for 2-PO-36.3 that pump flow rates for the 1A low head Safety Injection Pump of 1500 gpm and 2500 gpm were not used as data points as required
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by step 4.1.5 which reads in part, "... collect data...for 1500 gpm, 2500gpm, run out flow and any other flow condition desired by the Test Engineer...."
However, the data points which were tahen by the Test Engineer were adequate to verify the pump head performance curve. The inspectors concern was that the procedure had received the appropriate levels of review without this discrepancy being discovered and properly documented.
In view of the fact that objective evidence of review exists in the form of (1) a memorandum (reviewed and approved by SNSOC and JTG on 3/17/80) regarding " Supplemental Information to QA Records" in addition to (2) 9 cases of procedure deviations which were reviewed and approved, and that (3) the inspector found no review discrepancies in the remaining procedures, it is concluded that this item is not a noncompliance as it is unique and not indicative of the licensee's present program of review and evaluation.
b.
Completed preoperational test procedure results evaluations were verified for the following procedures:
(1)
2-PC-3.3, Revision 0, dated 3-30-78, Fire Protection System Smoke Detectors, results approved 3-5-80 (2)
2-P0-32, Revision 0, dated 8-9-79, Boric Acid Heat Tracing, results approved 3-5-80.
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(3)
2-PO-54, Revision 0, dated 12-6-78, Turbine Bunback Control Systee, results approved 2-25-80
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2-P0-59, Revision 0, dated 1-8-79, Condensor Steam Dump control, results approved 2-6-80 6.
Licensee Event Report Review (Unit 1)
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.
Additionally, 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.
LER 80-03, dated February 19, 1980, Nuclear instrumentation channel a.
N-44 inoperable, b.
LER 80-06, dated February 7, 1980, Casing cooling valve MOV-RS-100A found inoperable.
LER 80-09, dated February 6,1980, Six main steam safety valves lift c.
setpoints found to be incorrect.
d.
LER 80-11, dated February 13, 1980, Ventilation stack radiation monitors inoperable.
LER 80-13, dated February 11, 1980, Auxiliary feedwater pump P-2 e.
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LER 80-15, dated February 7, 1980, Average coolant temperature less than limits with the reactor critical.
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LER 80-20, dated February 28, 1980, One train of casing cooling inoper-
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able during a design change.
h.
LER 80-21*, dated February 28, 1980, Emergency diesel IH inoperable.
l The inspector reviewed deviation report (DR) 80-83 and maintenance report (MR) N1-8002021420 relating to the diesel false start on February 2,1980. Further discussions with licensee engineerinE staff l
revealed that the vendor has been contacted for evaluation of the
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crankcase pressure trip setpoint and time delay. The engineering staff felt certain that several earlier diesel trips (see LERs 79-88 and 79-131) occurred due to this fault, but were notidentified because
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operators reset the diesel after it tripped. The licensee plans to conduct tests on the IH diesel, or the 2J diesel which shows identical j
characteristics, to evaluate the crankcase pressure transient and
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assess whether this transient is due to normal wear oh, the diesels or otherwise. The results of this evaluation shall be fdllowed up in future inspections (338/80-13-01 and 339/80-14-01). The issue of emergency diesel repairs is discussed further below, in paragraph 61.
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I-4-1.
LER 80-22*, dated March 3, 1980, Quadrant power tilt greater than 1.02.
The inspector reviewed procedure 1-PT-22.2 "Incore vs. Excore Axial Offset" and discussed the event with the licensee's engineers.
During conduct of steps 4.3 through 4.5 of 1-PT-22.2 the quadrant power tilt, as indicated by the excore nuclear instruments, indicated greater than 2% tilt. Two situations exist which appear to contribute to the indication of a tilt during this procedure:
(1) Nuclear instrument N-44, one of four channels of power range instrumentation, has new detectors which will demonstrate a more rapid sensitivity change than the older detectors installed in the other three quadrants of the core. This could result in tilt indications between channel calibrations.
(2) A 0.8% quadrant power tilt skewed to the southwest quadrant does exist at 100% power.
(See LER 80-17).
Since this procedure results in recalibration of the excore detectors, this problem appears to pose no significance as long as the results of the procedure shows that the actual quadrant tilt was less than 1.02, as it was in this case.
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LER 80-23, dated March 10, 1980, Containment Hi-Hi pressure bistable inoperable on one channel.
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LER 80-25, dated March 14, 1980, Steam line pressure bistable setpoint inaccurate on one channel.
1.
LER 80-32*, dated March 14, 1980, Emergency diesel IH inoperable. The inspector reviewed DRs30-101, 80-137,80-145, 80-149, and 80-152 and MR N1-8002150730 relating to 5 diesel false starts. Discussions with licensee engineering staff and maintenance personnel on this report and the report previously described in paragraph 6h, indicated that the overspeed problem could be directly related to a servomotor malfunc-tion. The servomotor which was removed from the IH diesel contained debris which jammed the operating piston and caused the fuel racks to remain full open after starting air was removed. This servomotor, a Woodward Governor Co. Part No. 013145, was replaced with a similar design Woodward servomotor from a unit 2 diesel (part no.
8901-037).
This replacement unit is not identical to the original servomotor, as indicated in the LER, and because the vendor has indicated that the original servomotor is obsolete, raises the concern of possible generic problems with the model on other Colt Industry diesels in use at nuclear plants. This issue is being pursued by the NRC Vendor Inspection l
program and will be followed up by NRC (338/80-13-02).
Several crankcase pressure trip malfunctions are also. involved in this LER and are identical to the problem presented in LERL80-21 discussed in paragraph 6h above. Licensee engineering personnel' indicated that they are working with the vendor, Colt Industries, to resolve the
crankcase pressure setpoint and delay time settings to prevent diesel
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-5-g false starts during manual operation. The licensee has proven the reliability of these diesels to start in the manual mode at this time, and since this trip is blocked during automatic start it does not appear to pose a safety problem. Resolution of this crankcase pressure trip shall be followed up as item (338/79-50-02), which was previously opened when the cause of the diesel trips was unidentified.
No items of noncompliance or deviations were identified in the licensee event reports.
7.
Previously Identified Items of Concern a.
(Closed) 339/78-29-03 - This item related to the preservice inspection of the diesel auxillaries. The diesels have been successfully placed in service and this item is closed.
b.
(Closed) 339/78-29-06 - This item identified the need for a preservice valve lineup of the diesel lube oil and cooling water systems. The diesels have been successfully placed in operation and this item is closed.
c.
(Closed) 339/79-15-01 - This item identified the need to verify the manual operating features of the fuel manipulator crane. The licensee completed this item during the preoperational test 2-PO-30.4 on the manipulator crane. This item is closed. The licensee intends to repeat the preoperational test of the manipulator crane, including the manual operating features, just prior to fuel load.
d.
(Closed) 339/79-25-08 - Results of test 2-PO-39.1 on the residual heat removal system indicate that RHR pump 2-RH-P-1B successfully met its acceptance criteria. This item is closed.
e.
(Closed) 339/79-45-01 - The licensee has added to the file of results of preoperational test 2-PO-42.2 of the RCS loop isolation valves a letter from Westinghouse stating that the test results are satisfactory.
The licensee properly reviewed and approved this addition on 9/13/79 and this item is closed.
f.
(Closed) 339/79-50-01 - The licensee has reviewed documentation problems and properly supplemented the test record of preoperational test 2-PO-65.2 on RPS and ESF response times. This item is closed.
g.
(Open) 339/79-50-02 - The licensee has determined that the results of test 2-PO-65.2 can not be evaluated to clearly' meet the acceptance criteria. This test relating to reactor protection response time will be rerun before fuel loading. The results of the retest will be reviewed during a future inspection.
h.
(0 pen) 339/79-25-07 - Boric acid pump 1-CH-P-2D is out; of service for repairs. When repairs are complete a special procedu'W vill be performed r
to preoperationally teet thir pump. ' The results of this work will be reviewed during a future inspection.
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(Closed) 339/79-11-13 - Administrative Procedure ADMIN-10 was revised on September 5,1979, to include Unit 2 instrumentation in the instrument calibration program. A sampling review of the instruments and licensee documentation indicates the program is in effect.
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(Closed) 339/79-11-24 - The Nuclear Power Station Quality Assurance Manual section 12 paragraph 2.2 was corrected by revision 7 dated June 12, 1979. A sampling review of measuring test equipment by the inspector indicated that instruments used to verify limiting conditions for operations are in the calibration program.
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The following four items relating to Unit 2 snubbers will remain open to be reviewed during a future inspection:
(1)
(0 pen) 339/79-06-02 - The NRC staff will review the need to periodically functionally test snubbers associated with the reactor coolant system.
(2)
(0 pen) 339/79-06-03 - The documentation of an inspection of snubbers and hangers after hot functional testing will be reviewed during a future inspection.
(3)
(0 pen) 339/79-26-01 - The records of the resolution of deficiencies observed on snubbers and hangers during hot functional will be reviewed during a future inspection.
(4)
(Open) 339/79-26-02 - The licensee plans to visually inspect all Unit 2 snubbers at 550*F RCS temperature prior to criticality.
The results of this work will be reviewed during during a future inspection.
1.
The following two items concerned Emergency Procedures for Unit 1:
(1)
(Closed) 338/80-10-01 - Emergency Procedures 1-EP-2, 1-EP-3, 1-EP-4, and 1-EP-5 were revised on March 13, 1980 to change the reactor coolant pump (RCP) manual trip setpoint to 1485 psig reactor coolant system pressure. This change also prohibited RCP restart until safety injection reset criteria are met and manage-ment evaluation directs such action. A similar item (339/80-08-01)
for Unit 2 remains open pending inspector review of the final revised procedures.
(2)
(Closed) 338/79-09-02 - Emergency Procedures 1-EP-1 through 1,EP-5 have been revised to assure adequate verification of automatic actions, except as specifically noted in IE Report 338/80-10 and 339/80-08, as required by the approved Westinghouse guidelines.
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