IR 05000338/1979038
| ML19260A903 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/25/1979 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19260A896 | List: |
| References | |
| 50-338-79-38, 50-339-79-44, NUDOCS 7912060114 | |
| Download: ML19260A903 (7) | |
Text
4'pnarg 'o UNITED STATES
!
g NUCLEAR REGULATORY COMMISSION n
A E
REGION 11
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101 MARIETTA ST., N.W., SUITE 3100 o
ATLANTA, GEORGIA 30303
.....
Report Nos. 50-338/79-38 and 50-339/79-44 Licensee: Virginia Elect.ric and Power Company P. O. Box 26666 Richmond, Virginia 23261 Facility Name: North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and CPPR-78 Inspected at.'ionh Anna yte ear M,i ral, Virginia Inspected by:
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M. S. Ki 'd lleiddmc4pectog Dat/ Signed
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Approved by:
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P. J g e 1 E e g Chief, RONS Branch Day'e signed S
SUMMARY Inspected on September 4-14 and 20-25, 1979.
Unit 1 Areas Inspected This routine inspection by the resident inspector involved sixteen hours onsite in the areas of previously identified open items, licensee event reports, plant tours, new fuel storage, and licensee inspection of certain induction disc relays.
Unit 1 Findings No items of noncompliance or deviations were identified.
Unit 2 Areas Inspected This routine inspection by the resident inspector involved eighteen hours onsite in the areas of previously identified open items, reports per 10 CFR 50.55(3)
and 10 CFR 21, startup testing program, plant tours, and fuel storage:
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Unit 2 Findings No items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager J. R. Harper, Instrument Supervisor
- S. L. Harvey, Operating Supervisor
- W. R. Madison, NRC Coordinator
- E. R. Smith, Superintendent of Technical Services J. P. Smith, Reactor Engineering D. E. Thomas, Electrical Maintenance Supervisor Other Organizations Stone and Webster Engineering Corporation (S&W)
L. L. Evans, Unit 2 Area Supervisor
- Attended one or more exit interviews.
2.
Exit Interview The inspection scope and findings were summarized on September 6 and 14, 1979, for those persons indicated in Paragraph I above.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Plant Status Unit 1 experienced a turbine trip / reactor trip on September 25, 1979 (see IE Report 50-338/79-39). VEPC0 initiated a shutdown on that date to begin a refueling outage which had been scheduled to begin October 5, 1979. The outage was expected to last about twelve weeks.
Unit 2 was in the final stages of preoperational testing at the conclusion of the inspection period.
6.
Followup on Previously Identified Open Items Excessive Containment Radiation Levels (339/78-12-02)
a.
Licensee letter 300A of January 31, 1979, included as an attachment a description of the supplemental neutron shielding to be installed in Unit 2 containment to reduce radiation levels to FSAR values (problem 1510 087
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reported April 28, 1978 per 10 CFR 50.55(e)).
The supplemental shielding consists of a borated silicon rubber material encased in stainless steel, placed in a collar arrangement around the upper end of the reactor vessel, in a saddle shape over the inlet and outlet nozzles, and in blocks over the reactor vessel support feet.
The following documentation was reviewed by the inspector:
(1) Completed Engineering and Design Coordination Reports (E&DCR)
P-2502, P-2530 through P-2530D, 25778, 25779, 25785, 40189, 40191 through 40191B and 40195.
(2) Nonconformance and Disposition reports 3692, 3703, 3709, and 3720.
(3) Unit 2 Drawings 12050-FV-237A, B, and C.
The Unit 2 reactor cavity was not accessible, but the Unit I components to be installed, which are essentially identical to those of Unit 2, were observed. These compared favorably with the drawings and E&DCR's listed above.
In addition to the vessel shielding, Permali shielding was to be installed in crane wall openings between "A" steam generator cubible and the area of the personnel hatch on the operating deck of containment.
Visual observation confirmed this installation to be complete.
The inspector had no further questions concerning installation of shielding, but noted that confirmatory measurements of its adequacy will be taken during power ascension testing as defined by Table 14.1-2 of the FSAR.
Item 339/18-12-02 will remain open pending completion of the startup surveys.
VEPC0 confirmed by letter no.
262A on September 14, 1979, that the same shielding will be installed in Unit I during the refueling outage scheduled to begin in early October.
Item 338/78-14-04 also remains open.
b.
Moderator Temperature Coefficient Measurement (338/78-27-02)
As noted in IE Report 338/78-27, the core moderator temperature co-efficient (M70) was initially positive, requiring imposition of rod withdrawal limits until the MTC was determined to be negative by further testing (LER 78-26/99X-1).
In accordance with Technical Specification 4.1.1, MTC was measured May 22, 1979 by surveillance procedure 1-PT-13, and found to be - 27.1 percent milli rho (PCM) per degree Farenheit. This is within the limit set by the specification, thus this open item is closed.
Reactor Coolant Flow Calibration Procedures (38/78-40-01)
c.
As discussed in IE Report 50-338/78-40 and LER 78-091, a procedural error resulted in an incorrect, nonconservative setpoint for one flow channel. Of the nine (9) calibration procedures for this function, all of those containing a similar error were revised June 18, 1979 to incorporate correct setpoints for the individual channels.
Four such procedures were verified to have been revised on the dates above.
This item is closed.
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d.
Lack of Independence Control Room Air Conditioning Systems (339/79-12-02)
This Unit 2 problem was reported to NRC per 10 CFR 50.55(e) and 10 CFR 21 via letter 037 dated January 18, 1979. Air handling Units 2-HV-AC-06 and 2-HV-AC-09, which were powered from the "2J" Emergency Bus had their piping connected so as to normally receive chilled water from chiller 2-HV-E-04B, which was powered form the "2J" emergency bus.
Thus a failure of either emergency power source would dit.ahle both of the redundant subsystems for control room and emergency sw tchgear room air conditioning.
The solution to this problem, given by E&DCR P-2536-2, was to inter-change the supply and return chilled water lines for the air handling units such that the normally aligned chiller would be powered from the same emergency bus as the air handling units being supplied. As reported in IE Report 50-339/79-21, the piping modifications were verified complete, but valve lineups had not been revised to reflect the desired alignments.
During the current inspection, the inspector reviewed 1-0P-21.6A, Valve Checkoff - Control and Relay Room Chilled Water, revised June 18, 1979 and confirmed that the new lineups had been incorporated. Additionally, it was observed that Unit 1 drawings FB-40C and FB-26A had been revised appropriately.
Item 339/79-12-02 is considered closed.
During the review of 1-0P-21.6A, three (3) typographical errors were observed in valve numbers, although the valve descriptions were cor-rect.
Station management stated that these would be connected and the checklist would be reviewed again to assure accuracy. There were no further questions by the inspector.
7.
Startup Testing Program - Unit 2 This inspection effort was a continuation of that documented in IE Report 50-339/79-46. The scope of the Unit 2 post fuel load testing program was reviewed to ascertain whether FSAR commitments would be met.
Test objec-tives in individual startup test procedures, most of which were listed in IE Report 50-339/79-15, and 2-SU-1, " Nuclear Steam Supply System Startup Sequence and Administration," were reviewed and compared to the commitments of FSAR Table 14.1.-2, as modified by VEPC0 letters no.
06F of February 5, 1979; no.
150 of March 16, 1979; no.
240 of April 10, 1979; and no.
314 of May 2, 1979.
One discrepancy was noted.
Item II.2 of : Table 14.1-2 states that a power coefficient measurement test would be conducted at 65 percent power or greater, but this is not reflected in the letters referred to above. Station personnel stated that the FSAR would be amended to clarify that such a test would not be conducted.
Open item 339/79-44-01 is identified for followup of this matter.
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8.
General Electric Induction Disc Relays In September of 1978 General Electric (GE) notified NRC of a potential problem with certain induction disc type protective relays which might be in use in nuclear power plants. The problem involved use of petroleum jelly on relay time dials which could migrate to relay backstops, resulting in higher than normal pickup values on the relay discs. Discussions with licensee personnel and review of documentation revealed that a letter frorr GE on the subject had been received September 13, 1978 and appropriate supervisors vithin VEPC0 had been notified on September 19, 1978.
According to VEPC0 Automation and Control (A&C) personnel, no problems have been experienced on Units 1 or 2.
All relays have been tested on Unit 2 and those in Unit I will be rechecked for normal operation during the Unit I refueling outage per station electrical preventive maintenance procedures.
If " sticky" backstops are found, a cleaning procedure is available for use.
The inspector had no further questions on this subject.
9.
Licensee Event Reports - Units 1 and 2 The following events or problems were reported per Unit 1 Technical Specifications and/or 10 CFR 50.55(e) and 10 CFR 21 during the inspection period:
Emergency Diesel Generator Fan Cooling System a.
On September 10, 1979, licensee personnel reported per 10 CFR 50.55(e)
and 10 CFR 21 that a potential problem existed on the diesel generator-(DG) on cooling system for Unit 2.
The fan blades do not appear to have the proper pitch setting to provide necessary cooling for the wide range of temperatures of cooling air that can occur during the year.
It was noted that a meeting with Colt Industries, the manufac-turer, was scheduled to obtain more information on this problem. A similar problem for the Unit 1 DG's was reported September 11, 1979 as a prompt (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) licensee event report (LER).
This problem will be reviewed in more detail (open items 338/79-38-01 and 339/79-44-02).
Discussions with licensee personnel revealed that daily temperature extremes in spring and fall would represent the worst condition regarding fan blade pitch settings. Documentation available in system release package 1-22-12 confirmed that the Unit I fan blades were set for summer conditions, which can be expected to exist at least until the refueling outage, b.
Refueling Water Storage Tank Level-Unit 1 On September 11, 1979, station management related that the refueling water storage tank (RWST) level required by Unit 1 Technical Specification 3.5.5 might not be compatible with the safety analysis report. The maximum contained volume permitted in the RWST, 464,000 gallons, might not provide enough water in the containment following an accident because not all of the volume is usable and emergency
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procedure 1-EP-2, " loss of Reactor Coolant Accident," terminated suction from the RWST prior to expanding all the usable contents.
After further evaluation management reported on September 12 that two prompt LER's had been identified.
(1) The curve for RWST level,1-SC-5.2, was found to be based on usable volume rather than contained, thus the RWST level had been above the limits of Technical Specification 3.5.5 since startup.
The curve was corrected and the RWST was drained to bring level within limits within the time frame required by the associated Action Statement.
(2) Procedure 1-EP-2 was found to be in error in that not enough water would have been expended from the RWST prior to shutdown of quench spray pumps and initiation of recirculation flow using the containment sump. A standing order was iss!ted to modify 1-EP-2 until it could be formally revised to require longer operation of the quench spary pumps to obtain the necessary amount of water in containment.
The inspector had no further questions at that time, but will review the event in more detail following submittal of the four-teen day written LER's (open item 338/79-38-02).
c.
Impact of Non Safety Equipment on Protective Functions Units 1 and 2 On September 19, 1979, VEPCO reported that Westinghouse had informed licensee personnel that it could be possible for certain control systems, previously assumed to remain "as is", to fail when subjected to an adverse environment caused by high energy line breaks, thereby causing an impact on present safety analysis. This was reported as a prompt LER per Unit 1 Technical Specifications and subsequently per 10 CFR 50.55(e) and 10 CFR 21 for Unit 2.
Initial evaluations by licen-see personnel and the inspector revealed no areas of the present safety analysis which would be adversely impacted, but this matter will be reviewed in more detail following submittal of the LER and reports per 10 CFR 21 and 10 CFR 50.55(e) (open Items 338/79-38-03 and 339/79-44-03).
10.
Fuel Storage - Spent Fuel Pit Amendment 14 to the Unit 1 Technical Specifications, issued August 17, 1979, authorized the use of high density storage racks in the spent fuel pit. On September 11, 1979, the inspector observed the new racks as in-stalled. No damage was observed and and general cleanliness and house-keeping was good.
It was confirmed tht 966 storage positions were installed in agreement with the revised Technical Specification 5.6.3.
It was also observed that new fuel for Units 1 and 2 were stored with a spacing of 28 inches between elements as required by Specification 5.6.1 when stored dry.
Genereal area housekeepinE was in order. Security provisions were at least equivalent to those required by the Station Security Plan and special nuclear material license.
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11.
Plant Tours Tours of selected plant areas were conducted on September 7, 11 and 21, 1979. The following items, as available were observed:
a.
Fire Equipment Operability and evidence of periodic inspection of fire suppression equipment.
b.
Housekeeping Minimal accumulations of debris and maintenance of required cleanliness levels in systems under or following testing.
c.
Equipment Preservation Maintenance of special preservative measures for installed equipment as applicable.
d.
Component Tagging 1mplementation and observance of equipment tagging for safety or equipment protection.
e.
Communication Effectiveness of public address system in all areas toured.
f.
Equipment Controls Effectiveness of jurisdictional controls in precluding unauthorized work on systems turned over for initial operations or preoperational testing.
g.
Foreign Material Exclusion Maintenance of controls to assure systems which have been cleaned and flushed are not reopened to admit foreign material.
h.
Security Implementation of security provisions for both Units.
Within the above areas, no items of noncompliance or deviations were observed when compared to the applicable station programs and procedures.
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