IR 05000338/1979052
| ML19294B701 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/11/1980 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19294B691 | List: |
| References | |
| 50-338-79-52, 50-339-79-58, NUDOCS 8003050359 | |
| Download: ML19294B701 (9) | |
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
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,f 101 MARIETTA ST N.W., SUITE 3100
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ATLANTA, GEORGIA 30303 Report Nos. 50-338/79-52 and 50-339/79-58 Licensee: Virginia Electric 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 Inspection at North te Mineral, Virginia
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Inspector:
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Date Signed Approved by:
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P. J. Kylloggf Sectinn M af r RONS Branch Datt Signed f('
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SUMMARY Inspection on December 3-21, 1979 Unit 1 Areas Inspected
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This routine. inspection by the resident inspector involved 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> onsite in the areas of previously identified open items, general employee training, emer-gency procedure training, plant tours, licensee event reports, and plant status relative to startup from the refueling outage.
Unit 1 Findings: No items of noncompliance or deviations were identified.
Unit 2 Areas Inspected This routine inspection by the resident inspector involved 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> onsite in the areas of previously identified open items, plant tours, reports per 10 CFR 21 and 10 CFR 50.55(e), and Unit readiness for fuel loading.
Unit 2 Find ngs: Of the three areas inspected, no items of noncompliance or deviations were identified in two areas; one item of noncompliance was found in one area of (Infraction - Failure to review potentially reportable matters for applicability - paragraph 6).
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DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager S. L. Harvey, Operating Supervisor
- G. A. Karm, Engineering Supervisor
- J. D. Kellams, Superintendent of Operations
- W. R. Madison, NRC Coordinator
- J. L. Perkins D. E. Thomas, Electrical Maintenance Supervisor Other licensee employees contacted included two Health Physics, four operators, and two training department staff members.
Other Organizations Stone and Webster Engineering Corporation (S&W)
R. F. Ancerson, Advisory Engineering R. J. Daly, Lead Advisory Engineer L. L. Evans, Area Superintendent - Units I and 2
- Attended one or more exit interviews
- Contacced by telephone 2.
Exit Interviews The inspection scope and findings were summarized on December 7, 12, and 21, 1979, for those persons indicated in Paragraph I above. The apparent Infraction defined in Paragraph 6 was discussed December 7 and 12, 1979.
On the latter date, station management stated that the Station Manager would assume the duties of Proj ect Engineer in reviewing matters for reportability on Unit 2 in order to assure more timely reviews.
A. P.
Tattersall, Resident Inspector Trainee, was introduced to station management on December 17, 1979.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspectio.
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5.
Followup on Previously Identified Open Items Diesel Generator Muffler Seismic Supports (339/79-12-04)
a.
Efforts related to the last inspection of this problem were documented in IE Report 50-339/79-41, Details paragraph 8.b.
Questions regarding completion status of the modified line supports resulted in further review by Stone and Webster (S&W).
This review determined that a temporary, vertical support on each exhaust line should be removed and grouting behind the anchor points for the vertical constraints for each exhaust line should be removed and replaced with shims. Preopera-tional Deficiency Reports (PDR) 2-139 and 2-140 were issued to control these two tasks.
PDR 2-140 was actually implemented by Engineering and Design Coordination Report (E&DCR) P-2594B-2 and Rework Control Form (RCF) H-1056.
The inspector had no further questions after review of E&DCR P-2594 and the documentation referred to above and inspection of the equipment as modified. Item 339/79-12-04 is closed.
Inspection of the Unit 1 modification is discussed in IE Report 50-338/
79-15.
b.
Excessive Containment Radiation Levels (338/78-14-04)
VEPCO 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 reported for Unit 1 per LER 78-33). VEPCO letter 262A of September 14, 1979, notified the NRC that the same shielding would be installed in Unit I during the current refueling outage. On December 3, 1979 the inspector visually confirmed installation of the shielding and compared it to the provisions of design change (DC)79-507, including Field Changes one through four and drawings 13075-FV-237 A,B, and C.
There were no questions on the installation. It was observed that a Special Radiation Work Permit was in effect for this installation effort and it was being monitored by Health Physics as required.
Item 338/78-14-04 remains open pending completion of radiation surveys at power after restart.
Inspection of the Unit 2 shielding installation was dicumented in IE Report 50-339/79-44.
Inside Recirculation Spray Motor Bearings (338/79-20-01)
c.
Unit 1 LER 79-43 discussed a potential problem with lower bearing tempera ture if the bearings and shafts were built to their most detrimental tolerances, and committed to replacement of the bearings during the first refueling outage. Results of special twelve hour runs on the Unit 1 motors were discussed in IE Report 50-338/79-20.
On December 19, 1979, the inspector reviewed results of pump tests following bearing replacement conducted by special test procedure 1-ST-12, " Recirculation Spray System Pump Special Test."
The procedure required testing on recirculation flow to the containment sump, with stable bearing temperatures for at least four hours and a total run of each pump of at least twenty-four hours.
Test data revealed that
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bearing temperatures stablized af ter five to six hours of operation and remained stable for the duration of each test. Certificates from General Electric Company regarding bearing replacements were also reviewed.
Following discussions on the above documentat. ion with station personnel, the inspector had no further questions.
Item 338/79-20-01 is closed. A similar item for Unit 2 was closed in IE Report 50-339/79-46.
d.
Spare Pipe Penetration Caps (338/79-09-01)
As noted in LER 78-122, the licensee had discovered that the outside end of spare penetrations had not been capped as required by the FSAR.
This problem also applied to Unit 2 and inspector closeout of it for that Unit was documented in IE Report 50-339/79-33. During the current inspection, the inspector reviewed completed design change (DC)79-510 and related documentation for Unit 1.
This documentation demonstrated placement of caps on all the spare penetration except for four which were to be used by virtue of other design changes. Certain documenta-tion associated with DC 79-510 was verified to be part of the package but was not reviewed in detail. These included completed E&DCR's,
Noncompliance and Disposition Reports, QC Inspection Reports, Hydro Test Reports, Material Test Reports, isometrics, weld data sheets, etc. Eleven penetrations were randomly selected for visual inspection to confirm completion, line sizes and capping method. No items of noncompliance or deviations were identified and item 338/79-09-01 is considered closed.
Foriegn Substance in Valve (338/79-20-03)
e.
As noted in IE Report 50-338/79-20 and LER 79-48, during local leak rate testing of Unit I containment isolation valves, casing cooling valve MOV-RS-100B (FSAR Figure 6.2-73) failed its test. Upon disas-sembly of the valve, licensee personnel discovered a wax-like substance on the valve seat and disc. The inspector observed the disassembled valve and discussed the finding with plant management.
Immediate corrective actions included replacement of the valve and successfully conducting a leak rate test on the replacement.
Station management stated that they were confident that this was the only valve involved because of the construction history of the system, inspection of other valves, and discussions with the individual who apparently placed the foreign material in the valve. Following review of LER 79-48 and the circumstances surrounding this valve, the inspector had no further questions. Item 338/79-20-03 is closed.
f.
Refueling Water Storage Tank Level Errors (338/79-38-02)
As discussed in Unit 1 LER's79-111 and 79-112, the refueling water storage tank (RWST) level required by Unit 1 Technical Specification 3.5.5 was not 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 as accident because
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not all of the volume is usable and emergency procedure 1-EP-2, " Loss of Reactor Coolant Accident," terminated suction from the RWST prior to expending all the usable contents (LER 79-111). Additionally, the RWST level curve was in error (LER 79-112). As discussed in IE Report 50-338/79-38, a standing order was issued to assure use of all available water in the RWST and the level curve was corrected shortly after discovery of the errors.
During the current inspection, the inspector reviewed 1-EP-2, revised December 4, 1979, and confirmed that it had been revised to require continued operation of the quench spray pumps until all RWST water had been used. VEPC0 submitted a request to change Technical Specifications (Change No. 22) by letter 746 of September 13, 1979, to correct the RWST volume required. Discussions with the Unit 1 Project Manager in NRR on December 21, 1979, revealed that the requested change would be approved. Item 338/79-38-02 is closed. Unit 2 RWST level requirements and emergency procedures will be reviewed at a later date (open item 339/79-58-01).
g.
Limitorque Valve Actuator Manual Use (338/79-01-09 and 339/79-01-04)
As noted in IE heports 50-338/79-01 and 50-339/79-01, several valves using operators of the types described in IE Circulars 78-16 had been identified in use in Units 1 and 2.
Station personnel decided to place caution signs on the operators, stating that if the valves must be operated manually, they should be operated electrically af terward to assure proper lug engagement. During the current inspection, the
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inspector was informed that all signs had been placed on the operators involved.
All valves wer
- atirmed as being labeled as described above in Unit 2 (nine) and six signs were verified to have been placed in Unit 1.
Items 338/79-01-09 and 339/79-01-04 are closed.
h.
High Energy Line Break Effect on Steam Generator Level Indications (338/79-31-02)
As discussed in IE Report 50-338/79-31 and Unit 1 LER 79-83, Westing-house discovered that steam generator level indication may be higher than actual because of reference leg heatup following a high energy line break in containment.
Operating personnel were informed by Standing Order No. 31 on June 25, 1979, of the problem with correction factors to be applied to indicated levels following a line break based on containment temperature. Evaluation by station personnel revealed that the assumptions of FSAR Table 15.1-2 are still met assuming worst case error in indicated level feeding the Lo-Lo Level trip circuitry; however, a setpoint change was being considered to raise the trip from five percent to fif teen percent for added conservatism.
A review of two Unit 1 instrument calibration procedures (ICP) for steam generator narrow range level channels, ICP-P-1-L-475 and 494, revealed that they had been revised July 27, 1979, to raise the Lo-Lo Level trips setpoint to fif teen percent. According to station personnel,
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all others had been similarly revised.
In an unrelated matter a request to change the Unit 1 Technical Specifications trip setpoint to fifteen percent was made October 13, 1978 by letter no. 565. A review of Unit 1 emergency,rocedures dealing with high energy line breaks, 1-EP-2 and 1-EP-3, revealed that they had been revised to include precautionary statements concerning possible errors in steam generator and pressurizer level indications.
In order to minimize heatup effects, the steam generator reference legs were insulated during the current shutdown by DC 79-S58. Although this DC package documentation was not entitely complete, a review of it on December 21, 1979 revealed that the insulation had been installed.
Licensee personnel were informed that this item (338/79-31-02) was considered closed for Unit 1.
Unit 2 item 339/79-39-03 ccmains open and will be reviewed at a iater date.
IE Bulletin 79-21 was issued to all licensees concerning the heatup effects on level indications.
Review of VEPCO's response to the Bulletin is the responsibility of other NRC staff members.
6.
Potentially Reportable Items - Unit 2 As discussed in IE Report Nos. 50-338/79-48 and 50-339/79-56, Unit 1 LER's79-143, 152, and 153 had not been reported to NRC as being applicable to Unit 2 as of November 30, 1979.
In that these problems had been reported on Unit I on November 9 and 16, the inspector questioned whether a review for Unit 2 had been and if so, whe the conclusions were. On December 4, 1979 it was learned during a telephone conversation with J. L. Perkins, Director of Quality Assurance, that the Unit 1 Deviation Reports (DR) which resulted in the three LER's had not been reviewed by Power Station Engineering (PSE). Apparently the DR's had not been received by PSE or had been misplaced.
Licensee management was informed that lack of timely reviews for Unit 2 appeared to be in noncompliance with Criteria XVI of Appendix B to 10 CFR 50 and Paragraph 6.1 of Section 5.13 of Virginia Electric and Power Company's (VEPCO) Quality Assurance Manual, Engineering and Construction, which requires that each suspected failure to comply with the Atomic Energy Act or with NRC rules, orders, or licensee, and each deviation be reported on Form #883.8C.
The VEPCO employee, VEPCO architect-engineer, constructor, NSS vendor or consultant who discovers the suspected failure or deviation shall immediately forward the information to the Project Engineer, or his designee, for completion of Part 1 of Form #883.8C.
Paragraph 6.2 of Section 5.13 requires that a review be completed within five days, or the item becomes automatically reportable. The inspector further stated that this appeared to be an Infraction (339/79-58-02) and was a repeat of the apparent Infraction described in IE Report No. 50-339/79-28 and 50-339/79-35.
Station Management stated that the Station Manager would be designated to act as the Project Engineer for Unit 2 reporting for those matters wherin the station had the necessary information.
The three problems involved were subsequently reported under the provisions of 10 CFR 50.55(e) and 10 CFR 21 as noted in paragraph 1.
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7.
General Employee Training On December 14, 1979, the inspector attended a refresher training session of " General Employee Training" as participant and observer. This re-a training is required annually for all station visitors in order that a no-escort-required security badge be maintained by station administrative procedure (ADM) 12.0, " Station Training". Lecture content was in accordance with ADM 12.0 and a written examination was administered. No items of noncompliance of deviations were observed.
8.
Emergency Operating Procedure Training On December 13, 1979, the inspector attended a training session for operators and members of the training staff on these reviaed emergency operating procedures:
1-EP-1, Reactor Trip 1-EP-2, Loss of reactPReactor Coolant Accident 1-EP-3, Loss of Secondary Coolant 1-EP-4, Steam Generator Tube Rupture 1-EP-5, Safety Injection These procedures had been revised to accomodate recent developments in NL policy on operations of reactor coolant pumps and safety injection systems during accident conditions and Westinghouse criteria for operation and shutdown of this equipment.
Related changes to abnormal procedures (AP) were also discussed. No items of noncompliance or deviations were observed.
9.
Plant Tours
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Tours of various plant areas were conducted on several dates during the inspection period. 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. Observations regarding certain areas were given to station management who acknowledged the inspector's comments.
c.
Equipment Preservation Maintenance of special preservative measures for installed equipment as applicable.
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d.
Component Tagging Implementation and observance of equipment tagging for safety or equipment protection.
e.
Communication Effectiveness of public system in all areas toured.
f.
Equipment Controls Effectiveness of jurisdictional controls in precluding unauthorized work on systems turned over for intial 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 program and procedures.
10.
Licensee Event Reports - Unit 1 and/or 2 The following events or problems were reported as prompt (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) LER's per Unit 1 Technical Specifications and/or under the provisions of 10 CFR 50.55(e) and 10 CFR 21 for Unit 2 during the inspection period:
a.
Fuel Rod Burst Calculations (Unit 2)
On December 7,1979, VEPCO reported that this Westinghouse identified problem was also applicable to Unit 2 (See paragraph 6). This will be reviewed in more detail (339/79-58-03).
b.
Noncompliance in Error Analysis for Negative Rate Trip (Unit 2)
This was also reported as being applicable to Unit 2 on December 7 and is the same problem related in Unit 1 LER 79-152 (open item 339/79-58-04).
c.
Improper Analysis of Reactivity Profile During Dropped Rod Events (Unit 2)
This was also reported as being applicable to Unit 2 on December 7 and is the same problem reported in Unit 1 LER 79-153 (open item 339/79-58-05).
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d.
Differential Pressure Switch Not Seismically Qaulified (Units 1 and 2)
On December 12, 1979, VEPCO reported a discovery by S&W during reviews required by IE Circular 78-08 that differential pressure switches used in controls of the condensing water for the Units 1 and 2 control room air conditions were not seismically qualified (Unit 1 LER 79-158). No immediate safety concern was evident, but this will be reviewed in more detail following receipt of the Unit 1 LER and Unit 2 report per 10 CFR 50.55(e) and 10 CFR 21 (open items 50-338/79-52-01 and 339/
79-S8-06).
Incorrect Valve Weights Used in Stress Analyses (Unit 1)
e.
On December 19, 1979, VEPCO reported that a review required by IE Bulletin 79-14 had led to the discovery of incorrect weight assumed for an eight inch Fisher butterfly valve used in the component cooling water system. Use of the correct weight resulted in calculated over-stress of a hanger.
On December 20, 1979 it was reported that a second valve similar to the first had been discovered. These will be reported as IIR 79-148.
Station management stated that necessary modifications would be made prior to Unit I startup from the refueling outage (open item 338/79-52-02).
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