IR 05000338/1980019
| ML19290G644 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/21/1980 |
| From: | Kellogg P, Kidd M, Tattersall A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19290G639 | List: |
| References | |
| 50-338-80-19, 50-339-80-20, NUDOCS 8012180208 | |
| Download: ML19290G644 (22) | |
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NUCLEAR REGULATORY COMMISSION pM' E REGION 11 Q
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Report Nos. 50-138/80-19 and 50-339/80-20 Licensee Virginia Electric and Power Company P. O. Box 26666 Richmond, Virginia 23261 Facility:
North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and NPF-7 Inspection at North na Site ne9r ineral, Virginia N
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pector Dit'e Sfgned Approved by:
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SUMMARY Inspection on April 14 through May 30, 1980 Unit 1 Arees Inspected This routine inspection by the resident inspectors involved 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> on site in the areas of previously identified enforcement, unresolved, and open items and licensee event reports and reports per 10 CFR 21.
Unit 1 Findings Within the two areas inspected, one apparent item of noncompliance was identified (Infraction-failure to conduct a surveillance test at the required frequency -
paragraph Sc).
Also, two apparent deviations from written commitments were identified (failure to provide instructions for re-establishing ventilation flow to the safeguards buildings following a seismic event paragraph 5b'11) and failure to administratively control SW valves to the recirculation air coolers -
paragraph 8m).
Unit 2 Areas Inspected This routine inspection by the resident inspectors involved 194 hours0.00225 days <br />0.0539 hours <br />3.207672e-4 weeks <br />7.3817e-5 months <br /> onsite in the areas of previously identified enforcement, unresolved, and open items; ini-tial fuel loading; pre-critical testing; procedures for operation of a unit when 8012180Qe$
-2-an emergency occurs on the companion unit; procedures for use of the public address system; st.artup test engineer qualifications and training; reports per 10 CFR 21, 10 CFR 50.55(e), and the Technical Specifications, and unit readiness for initial criticality.
Unit 2 Findings Within the nine areas inspected, one item of noncompliance was identified (Infraction-failure to utilize the Rework Control Program in implementing a design change to the control room air conditioning chillers paragraph Sb(17)).
Two apparent deviations frora written commitments u re also identified (Failure to complete modifications to the control room air conditioning chillers and their control panels prior to initial fuel loading paragraph 5b(17)); and failure to conduct indoctrination / training in quality assurance procedures for a startup tes; engineer-paragraph 11.
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DETAILS 1.
Persons Contacted Licensee Employees
- D. L. Benson, Superintendent of Technical Services R. A. Bergquist, Assistant Instrument Supervisor
- W. R. Cartwright, Station Manager J. R. Eastwood, Senior Engineering Technician
- E. W. Harrell, Superintendent of Maintenance
- R. T. Johnson, Construction QA Engineer
- J. D. Kellams, Superintendent of Operations
- W. R. Madison, NRC Coordinator D. L. Snodgrass, Assistant Instrument Supervisor J. E. Wroniewiez, Construction QA Engineer Other licensee employees contacted included three technicians, four operators, a mechanic and office personnel.
Other Organizations Stone and Webster Engineering Corporation (S&W)
M. C Reynolds, Senior Advisory Engineer A. W. Rychalsly, Advisory Engineer E. W. Spurell, Site Engineering Office Babcox and Wilcox Four Startup Test Engineers
- Attended one or more exit interviews 2.
Exit Interview The inspection scope and findings were summarized on April 18 and 24 and May 9 and 30, 1980, for those persons indicated in Paragraph 1 above. A11 findings presented in these details were discussed.
3.
Licensee Action on Previous Inspection Findings (Closed) Deviation 339/79-35-01:
Improperly Sized Orifices Corrective measures described in the licensee's response of September 10, 1979, were verified to be complete by review of Engineering and Design Coordination Report (E&DCR) P-688A-2, and QC Inspection Reports 2-RS-24 and 2-RS-25, dated June 13 and 14, 1979. These documents verified placement of full base flanges in the pump discharge lines as required. As discussed La the licensee's response to this deviation, the licensee's report per 10 CFR 21 and 10 CFR 50.55(e) of June 18, 1979, and IE Peport No. 50-339/79-35, Details paragraph 9a, the improper orifice, sized for 3,000 gpm flow for
-2-each outside recirculation spray pump, was an interim measure na Unit 1.
This was superceded by installation of the casing cooling system to over-come a net positive suction head problem for the spray pumps. The licensee's response states that the E&DCR for Unit I which replaced the 3,000 gpm orifices with full bore flanges after casing cooling installation was not recognized as being relevant to unit 2 in that the interim modifications to Unit I had never been implemented for Unit 2.
Based on information previously reviewed concerning the E&DCR systems and discussions with licensee personnel, the inspector concluded that this omission appeared to be an icolated event.
This matter is closed.
(open) Infraction 338/79-39-06:
Cnanges to Equipment as described in the FSAR.
A portion of the corrective and preventive measures described in VEPCO's responses of November 29, 1979 and January 14, 1980, included training of operations and maintenance supervisors in the requirements of 10 CFR 50.59 and the Nuclear Power Station Quality Assurance Manual regarding reviews and evaluations required prior to changes in plant equipment and procedures.
The inspector reviewed documentation of training held on various dates through January 31, 1980 and ascertained that all supervisors had been instructed as committed. Lesson material was reviewed and discussed and one training session was attended on January 17, 1980 The inspector had no comments or questions on this facet of preventive measures but noted that other measures bad not been reviewed. This matter remains open.
(0 pen) Unresolved Items 338/79-45-04 and 339/79-54-06: Siesmic Qualitica-tions of Safeguards Building Exhaust Ventilation System. Interim hardware and procedural changes have been implemented.
Evaluation to determine if further changes are necessary is ongoing (Details paragraph Sb(11)).
4.
Unresolved Items No unresolved items were identified during this inspection.
5.
Unit 2 Status-Readiness For Operation IE Report 50-339/80-17, Details paragraph 7 delineated activities to be completed and problems to be resolved prior to Unit 2 proceeding with fuel loading, filling and venting the reactor coolant system, initial criticality, etc.
That report discussed the resolution of items required for fuel loading and two unit operation in modes 4 or above. Items cequired to be completed or othe rwise resolved prior to other milestones are discussed below. References are given to Section 7 of the Unit 2 Appendix A Technical Specifications which incorporated the problem areas identified by IE at the time of operating license issuance for Unit 2.
Items to be resolved prior to filling and venting the reactor coolant a.
system af ter fuel loading (technical specification section 7.2)
(1) Completion of Calibration and Checkout of the Overpressure Protec-tion System, Including the High Pressure Nitrogen System for the Power Operated Relief Valves (339/80-17-01
-3-The status of completion and checkout of the overpressure pro-tection system was discussed in IE Report 50-339/80-17 Details paragraph Sa.
At the completion of that inspection, certain calibrations and func tional checks were to be conducted and questions remained regarding power operated relief valve (PORV)
stroke time acceptable under dynamic conditions.
During the current inspection, the following completed procedures, used to calibrate and functionally test the overpressure protection system were reviewed:
-ICP-P-2-P-402-1, " Reactor Coolant System NDT Overpressurization Protection,"
-ICP-P-2-P-403, " Reactor Coolant System Pressure-Wide and Narrow Range Protection Channel IV"
-ICP-P-2-T-410/420f430, "Wioe Range Temperature (Cold Leg) Pro-tection Channel II"
-2-PT-44.4.1,
"Overpressurization Protection Instrumentation Functional Test with PORV Operability Not Required"
-2-ST-21, "NDT Overpressure Protection Functional Check Circuitry" Review of these completed procedures revealed that all inputs to the system including system logic had been calibrated or func-tionally tested as of April 25, 1980.
Regarding valve stroke time, stroking of the two P(RV's under static conditions revealed them to open in 1.9 second; as compared to the Westinghouse limit of 2.6 reconds.
The inspector quest-ioned whether the valves would be aroked under dynamic conditions representing operating conditions to assure acceptable opening time.
Station management's position was that, with flow ender the seat of this type of glG valve, dynamic strok e time would be equal or less than static time. To assure valve orientation was such that flow was under the seat, both PCV-24550 and PCV-2456 were radiographed on May 10, 1980.
Review of the radiographs resulted in concurrence by the inspector that flow is under tb-seats. There were no further questions on this matter, and item 339/80-17-01 is considered closed.
Based on the inspection efforts documenteri herein and in IE Report 50-339/80 17, Operating License NPF-7 condition 2D(5)o appears to have been satisfied for mode 2 operation.
(2) Replacement of Control Rod Guide Tube Support Pins (139/80-17-07)
This problem involved potential stress corrosion cracking of the upper internals control rod guide tube support pin as reported per 10 CFR 21 on March 18 (letter no. 237) and 10 CFR 50.55(e)
-4-(letter ao. 327) on April 10, 1980. The potential for failure existed because of solution heat treatment at temperatures below
~
1950 F.
As a conservative measure, the Unit 2 pins were placed by Westinghouse using procedures RBB-01-21-80, " Removal and Installation of Support Pins for Guide Tubes and Flow Columns,"
rod FCM-1-22-80, " Field Modification of Upper Internals Assembly tor Removal and Installation of Guide Tubes."
Portions of the changeout evolutions were witnessed on various dates, including tue following:
(a) Removal of the upper and middle sections of guide tubes from the upper support structure, and their storage and handling, (b) Hydro-lazer. leaning of sections o. the tubes, (c) Replacement of the middle guide tube sections, (d) Realignment of the guide tubes, (e) Replacement of the control rod drive, (f) Replacement of the control rod drive shafts into the upper guide structure.
Completed procedures, checklists, QA documentation, material certifications, etc.,
were reviewed.
It was noted that the
.r.eplacement pins had been solution annealed to temperatures ranging from 1945 F to 1967 1, which should preclude any cracking. The inspectors had no further questions on this mattcr for Unit 2.
Item 339/80-17-07 is closed.
b.
Items to be Resolved Prior to Initial Criticality (Technical Specifi-cation Secticn 7.3)
(1) Modifications to Incore Guide Tube Insulation Supports (339/78-24-02)
This problem, reported per 10 CFR 50.55(e) and 10 CFR 21, was le:t discussed in IE Report 50-33's/79-23, at which tir-VEPC0 committed to verifying that no in'.crferences existed between the modified supp' rts, the insulat.< u, and/or the incore guide tubes with the Unit at operating t<:mperature. This verification was accomplished by station engineers on May 21, 1980, as part of the visual inspections performed per 2-P0-48.1A, "Peactor Coolant and Associated System. Thermal Expansion and Restraint Inspection-Heatup Prior to Criticality." No interference were ob s e rved,
thus item 339/78-24-02 is considered close (2) Visual Verification of Hydraulic Snubber Operability at Operating Temperature 339/'9-26-02 This item involved an inspector's concern that all snubbers may not have been inspacted for adequate installation and operability at no rmal operating temperature during hot funct ional testing (HFT). In response to this concern, a two-foid inspection program was implemented during pre-critical testing. A program entitled
" HOT DAN H53 Inspection Program, North Anna Unit No. 2, May, 1980" was conducted May 21 and 22,1980. This included visual inspec-tions and measurement of hot piston stroke settings for those snubbers which were found to have deficiencies during HFT (about 48), snubbers added after HFI (About 19), and those that did not reach normal operating temperature during HFT (about 37, on the feedwater system). The snubbers on the feedwater lines will be checked after initial criticality when the secondary plant is heated up.
In addi: ion, test procedure 2-PO-48.1A, " Reactor Coolant and Associated Systems The rmal Expansion and Restraint Inspection-Heatup Prior to Criticality" was conducted May 16-21, 1980 at various temperature plateaus during heatup for precritical testing.
This test included a visual inspection of all snubbers on the reactor coolant and associated systems. All problems encountered were resolved by May 22, 1980.
The inspector had no questions following review of the completed inspections referred to above. This item is considered resolvea for purposes of initial criticality for Unit 2, but item 334/78-26-02 remains open pending completion of inspection of feedwater system snubbers during startup testing.
(3) Verify Adequate Seismic Stress Analysis of Safety Related Pic ng per IE Bulletin 79-07 (339/79-28-07)
VEPCO's response to this IEB was contained in letter no. 299 dated April 25, 1979. This response was reviewed by other NRC staff personnel and the inspector was informed May 28, 1980 by Region II that there were no questions on this response which would prohibit plant startup.
Item 339/79-28-07 is closed for criticality but remains open administratively for tracking purposes.
(4) Verification of Adequacy of Servicewater and Component Cooling Water Supports to Resist The rmal Loadings Over Service Water Temperature Range (339/78-36-01)
This item was inspected by another IE inspector, whose findings are documented in Report 50-339/80-23. It was found that c)rrec-tive measures were sufficient to assure safety of operation of the Uni (5) Determine if A Typical Weld Material Was Used in Reactor Vessel Seam Welds (IE Bulletin 78-12A)
The most recent VEPCO response to this IEB was letter no. 418B of Dece nber 26, 1979. The inspector was informed by Region II on May 28, 1980, that there were no questions on this response which would preclude plant startup.
(6) Erroneous Check Valve Weights (339/79-21-03) and IEB 79-04 This matter was closed in IE Report 50-339/80-23.
(7) DNBR Analysis Error For a Single Dropped Control Rod (339/79-28-01)
This problem was reported per 10 CFR 50.55(e) and 10 CFR 21 by letter 230 of April 6, 1979 and 230A of May 3, 1979. The latter letter defirred corrective action to include changing the Technical Specification setpoint for negative rate trips to assure that power overshoot would not occur.
During a subsequent meeting with NRC and V 'st inghouse, it was agreed that the Technical Specification change was not necessary and that administrative controls would suffice.
These controls, involving limiting cu atrol rod insertion limits when operating in the automatic rod control mode above 90% power, were defined by VEPCO's letter no.
973 of December 17, 1979. Operating procedure 2.1, " Unit Power Operation Mode 2 to Mode 1",
dated Jsnuary 30, 1980, and Station Curve 2-SC. 7, " Control Rod Insertion Limits versus Power Level",
dated March 30, 1980, were reviewed. The appropriate limitations were observed to be contained in these procedures, thus item 339/79-28-01 is considered closed.
(8) Resolution of Potential Degradation of Electrical Penetrations due to Overcurrent Conditions (339/79-46-01)
This item was initially reported in accordance with Part 21 ani Pat t 50.55(e) of 10 CFR and is identified as Open iter 09-46-01).
Engineering and Lerign Coordination Reports (E&DCR's; P-2688-2, P-2688A-2 thru P-2688K-2 and P-2716-2 and the results of the required tests were reviewed to determine if the hardware had been properly installed and the test results were satisf,ctory.
The results appeared satisfactory
.d this item (79-46-01) is considered closed.
(9) Diesel Generator Fan Cooling System (339/79-44-02)
VEPC0 submitted letters per 10 PR 21 and 10 CFR 50.55(e) on this problem on September 14, October 10, and December 14, 1979. The latter report stated that a modification to the fan drive system had been made to preclude the need for sersonal fan blade pitch adj ustments. It also stated that a 35 percent solution of ethylene
-7-glycol would be used during the winter months to prevent freezing and treated water would be used in summer months for maximum cooling.
Regarding the fan drive system, two additional bolts were added to the drive shaft coupling to allow use of increased fan blade pitch settings.
This was controlled by Unit 2 E&DCR P-2693-2, with revision and Rework Control Form (RCF) M2-1220.
Review of completed manintenance Reports (MR) N2-79-11271126and N2-79-11271125 revealed that the proper amount of ethylene glycol had been added to the cooling systems by December 8, 1979, which was within the time f rame recommended by Power Station Engineering.
To assure changeout of glycol and treated water at proper times of the year, the Mechanical Maintenance Prevention Mainter.ance Program was revised to schedule completion of the semi-annuals portions of MMP-P-EG-1 in May and October of each year.
Based on review of the completed E&DCR, MR's and review and discussion of the revised preventive maintenance program, the inspecto.:; had no further questions. Item 339/79-44-02 is closed.
A similar item for Unit 1 is also closed as discussed in paragraph 6.b of these Details.
(10) Cracked Dise For Containment Isolation Valve MOV-2380 (339/79-54-04)
This item was initially reported in accordance with Section 50.55e of 10CFI.
Corrective action was conducted under Rework Control Form (RCF) M2-1243 and 1246 and testing was completed using 2-PT-61.3 dated 2/15/80. Tnese documents were reviewed and appeared satisfactory. Item 339/7954-04 is considered closed.
(11) Seisimic Qualification of Ventilation Sjstems (339/79-54-06)
This prc' lem was reported per 10CFR 50.55(e) and 10CFR 21 for Unit 2 on October 31, 1979, with letters no. 907 of November 2, 1979, no. 907A af November 30, 1979, and no. 907b of January 18, 1980 veing submitted to define the problem and discuss prcy sed solutions. The letter of January 18, states that only the safe-guard area exhaust fans and auxiliary building central area exhaust fans need to be operable following a seismic event for those ventil-ation systems whose discharge ducts were not seismically qualified.
This is based upon the need for continued cooling of emergency equipment located in those areas. Additional corrective / preventive measures, including administrative and procedure changes to be employed were discussed in a supplementary letter to Unit 1 LER 79-149 dated December 28,1979 (serial no.1160).
-8-Letters nos. 1160 and 907B described a system design change involving installation of an exhaust port in the discharge duct-work for the central area exhaust fans which could be opened manually if the ductwork were to fail. Additionally, the ductwork from the fans out to the new exhaust port was reinforced to assure integrity during a seismic event. These physical changes were implemental in Unit 1 design changes DC-79-S87, consisting of E&DCR's OSD-0064-1 through revision 0064Q. On April 17 and 18, 1980, the inspector observed supports and the exhaust port as installed by E6DCR's OSD-0064E, G, L, P, and Q.
Installation appeared to be in conformance with the E&DCR's and accompanying drawings. The exhaust port was manually operable at that time.
"
m; sign change included provisions for automatic opening of the exhaust port upon a high discharge pressure in the ductwork.
Hardware associated with this function was not completely installed as of that iate.
Regarding administrative controls to assure continued equipment operability.
Letter no. 1160 of December 28, 1979 stated that emergency procedures would be revised to reflect the required operator actions to maintain ventilation flow during the postu-lated f ailure of non-seismic ductwork. During the inspection it was observed that abnormal operating procedure 1-AP-36, " Seismic Event," had been revised March 5,1980 to incorporate instructions in the event that the central area exhaust ductwork (vent stack A) collapsed. It was also observed, however, that there were no irntructions concerning the safeguards area exhausts (vent stack B). Station managen mt was informed April 24, 1980, that failure emergency procedures to provide instructions for starting, i
to rev ca the backup ventilation for the safeguards building if the normal exhaust systcm were Icst appeared to be a Deviation from a written commitment (338/80-19-01).
This comment was ad'nowledged and 1-AP-36 was revised April 25 and May 27, 1980 to provide such instructions and incorporate other comments by the inspector.
Letter no. 907B of January 1 1980 was an interim report per 10
,
CFR 50.55(e) and stated that evaluation of the ventilation systems was continuing to dete mine what additional action, if any, would be required. Duriag discussions on these continuing evaluations, the itspector questioned whether evaluations of adequacy of the safeguards buildings ventilation had been considered in non-seismic supply ductwork (FSAR Supplement 9.44).
He was informed that this had been considered and that updated information was expected from S&W in the near future, at which tiee an updated 50.55(e)
report would be submitted.
Unresolved items 338/79-45-04 and 339/79-54-06 remained open at the conclusion of the inspection perio (12)
Incorrect Seismic Response Spe:tren Curves Used in Stress Analyses (339/79-56-02)
Inis it am was reviewed by another inspector and closed i r. IE Repo:; 50-339/80-2'.
(13, CDA/S! Reset Function Errors (339/79-56-04)
This problem was initially reported in accordance with 10CFh 50.55(e) by letter, serial number 964 dated Nrvember 21, 1979.
The final report was made by letter, serial number 964A, dated December 14, 1979.
An additional equiprent problem involving TV-MS201A, B,
C was reported on April 11, 1980 and the final report, letter scrial number 412, was submitted on May 8, 1980.
E&DCR P-2702 was initiated to correct the reported problems w.;L the following equipment:
Control Room Dampers (SOV-HV-160-1,2; -161-1,2);
Iodine Filter Air Operated Dampers (S0V-HV-115 A2,B2);
Containment Air Recirculation Fans (2-HV-F-1A, IB);
Control Rod Drive Cooling Fans (2-IN-F-37A, B, C, D, E, F) ;
Service Water Radiation Monitor Pumps (2-SW-P-5, 6, 7, 8);
Recirculation Spray Pumps (2-RS-P-1A, IB, 2A, 2B);
Safeguards Area Air Operated Dampers (2-HV-228-1, 2, 3, 4);
Auxiliary Feedwater Pump Turbine Steam Supply Valves (2-TV-MS 211A, B).
This E&DCR, 2-ST-4 completed 3/19/80 and engineering study ER 80-49 dated 4/19/80 were reviewed and it appears that the designed and installed changes do insure that the control circuitry responds as described in the FSAR. The administrative controls or design features described in the above letters for control of Service Water to the containment Recirculation Air Coolers (2-MOV-210A,B, 214A,B); Condenser Air Ejecter Exhaust to Containment (2-TV-102-1, 103); and the Main Steam Isolation Valves (TV-MS201A, B,
C)
appear to provide adequate control and are considered adequate to close this item for initial criticality. This item (79-56-04)
will remain open however until the remaining hardware changes to TV-MS-201A, B, C and 2-TV-102-1,103 are completed an i tested.
(14) Nonconservatism in Error Analysis for Nuclear Instrumentation Negative Rate Trip for Dropped Control Rods (339/79-58-04)
This problem similar to that discussed in paragraph 5.b(7) of these Details, was also reported per 10 CFR 50.55(e) and 10 CFR 21 (letter no. 1117 of December 11, 1979). The corrective action was administrative limitations on control rod insection limits as discussed in paragraph 5.b(7). Item 339/79-58-04 is also considered closed.
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i-10-(15) Nonconservatism in The Method of Generating Power Distributions and Rod Wot ths for Dropped Rod Events (339/79-58-05)
This problem was also reported per 10 CFR 50.55(e) and 10 CFR 21 (letter no. 1116 of December 11, 1979) and corrective actions were the same as those discussed in paragraph 5.b(7) of these Details. Item 339/79-58-05 is considered closed.
(16) Seismic Analysis of As-Built Safety Related Piping Systems (IEB 79-14,339/80-09-01)
This subject was closed for purposes of initial criticality in IE Report 339/80-23.
(17) Eeismic Qualification of Control Room Air Conditioning Chiller Control Panels (339/80-05-01)
This problem was reported per Unit 1 LER 80-08 and per 10 CFR 50.55(e) for Unit 2 (letter no. 123 dated February 14, 1980). As discussed in the licensee's 50.55(e) report, a " shaker" test was conducted on one of the Unit 2 control panels. The test report (No. 15146) by Action Environmental Testing Corporation, dated February 8, 1980, concluded that the panels were seistically qualified if two changes were made: first, rubber mounting pads were to be replaced with metal shims to assure a rigid frame and secondly, a "Guardristor" relay would be electrically disconnected from the chiller control circuit. These changes were implemented
.for Unit 2 by E&DCh P-2/17-2, completed on April 28, 1980 for the electrical portion and April 1, 1980 for the mechanical portion.
Revies of the completed E&DCR and observation of the modified mounting pads resulted in no quertions by the inspector regarding hardware changes.
Item 339/80-05-01 is considered closed.
A similar Unit 1 item, 338/80-07-04, remains open.
Modifications to the Unit I chiller units were in progress at the conclusion of the inspection.
During the review of the Unit 2 problem, the inspector was informed that the mechanical rework had apparently been performed without a Rework Control Form, contrary to VEPCO Construction Quality Control Procedure 26.1, " Rework Control Program." The lack of an RCF wat documented via Nor. con fo rmance Report (NCR) 2-0009 on May 9, 1980. The NCR called for rework with QC present and was signed as being complete May 15, 1980.
Station manager.ent was informed that failure to utilize an RCF appeared to be in noncom-pliance with 10 CFR 50, Appendix B, Criterica V as implemented by VEPCO's Topical Report for Operation QA (VEP-3A) (Infraction -
339/80-20-01).
-11-VEPC0 letter no. 123 of February 14, 1980, stated it at the above modifications would be completed prior to inittsi fuel loading of Unit 2 (start date of April 12, 1980).
As noted above, the electrical modifications were not completed until April 28, 1980.
Station management was informed that this appeared to be a Devia-tion from a written commitment (339/80-20-02).
(18) Rerouting of Steam Gerie ra t o r Blowdown Monitor Em :ng Lines (339/79-54-02)
As discussed in IE Report 50-339/80-09, details paragraph 5.d, all Unit 2 protection system sensing lines had been found to be properly connected, but the sensing lines for the steam generator blowdown radiation monitors were not.
Preoperational Deficiency Reports (PDR) 2-161 through 164, 169, and 170 had been gene: 3ted to resolve discrepancies found.
During the cu rent inspection, the PDR's referenced above; E&DCR's PS-5322-2, PS-5322A-2, and PS-5322B-2; NCR 2-0007; and completed procedure IMP-C-PROC-23,
" Verification of Transmitter Process Connections," were reviewed.
The E&DCR's and NCR were used to control rework of the sensing lines and recheck instrumentation and alarms. The IMP was rerun to verif y proper sensing line connections following rework. The inspector had no further questions and item 339/79-54-02 is considered closed.
(19) Pressurizer Relief and Safety Valve Position Accoustical Monitoring System (339/80-17-02)
This subject was previously inspected and findings discussed in IE Report 50-339/80-17.
As of that inspection, control board indicators had not been installed, nor had all calibratior+ been completed. During the current inspection it was verified that the indicators had been installed and that calibration procedure ICP-2-VMS-01, " Valve Monitoring System," had been performed on April 22, 1980. There were no further questions following review of the completed ICP, thus this item is resolved for plant startup.
In that final setup of the accoustical monitoring system must be accomplished during power operation, item 339/80-17-02 remains open for tracking of that effort.
Installation and calibration of the control room indicators satisfies Operating License NPF-7 condition 2.D.(5).k for mode 2 operations.
(20) Completion of Installation and Checkout of High Range Radiation Monitors This short Term Lessons Learned requirement was previously inspected and closed in IE Report 50-339/60-1 (21) Safeguards Building Valve Pit Sump Pumps (339/80-17-08)
E&DCR's P-2687-2, P-2687A-2, P-2687B-2, Electrical Test Proce-dures (TEP-1, TEP-2) and Instrument Loop Functional Checkouts (TIP-3) were reviewed for the valve pit sump pumps (2-DA-P-6A,6B)
and the safeguards building sump pumps (2-DA-P-1A, IB). There were no questions in this area, thus item 339/80-17-08 is closed.
(22) Retest of Volume Control Tank Level Control Valve LCV 1 2115A (339/80-17-09)
This item was closed in IE Report 50-339/80-21.
(23) Completion of Reanalysis and Corrective Actions for Stresses in Safety Injection Lines for Temperatures Below 70 F (339/80-17-10)
This matter was reported per 10 CFR 21 and 50.55(e), with a final report submitted on April 8,1980 (letter no. 314).
A review of reanalyses and modifications was documented in IE Report 50-339/80-23, wherein the item was considered resolved for initial criticality with recalculation of pump nozzle loads and piping penetration loads to be completed in the near future. On May 29, 1980, the resident inspector was informed that the LHSI pump vendors, Ingersall Rand, had lowered the acceptable values previously used for pump nozzle loads, resulting in several (about 25) additional piping support modifications.
Design change documentation was in preparation for these changes at the conclusion of the inspection period.
Item 339/80-17-10 remains open. The added modifications are to be completed prior to Unit 2 initial criticality.
(24) Gaps in Grating of Containment Recirculation Sump Enclosures (339/80-17-11)
The sump grating was reinspected following repairs and appearc to conform ot the design as stated in FSAR Section 6.2.2.2.
Item 339/80-17-11 is considered closed.
c.
Items to be Completed Prior to Exceeding Zero Power Physics Testing Power Levels (1) Resolutien of Potential Nonconservative Feature of Westinghouse La rge Break LOCA-ECCS Evaluation Model and its Impact on Peak Clad Temperatures. (339/79-58-03)
VEPCO submitted supplemental information concerning this problem and requested a change to the Unit 2 Technical Specificationsvia Proposed Technical Specification change no. 27 (Units 1 and 2).
The proposed change was incorporated into the Technical Specifi-cations issued with Unit 2 operating License NPF-7.
Item 339/79-58-03 is considered close.
Procedure for Unit Operation During Opposite Unit Emergency In response to an inquiry f rom NRR, VEPCO committed via letter no. 313 dated April 3, 1980 to establish procedural guidance for operation of an unaffected reactor when the companion Unit is undergoing an emergency. This was to be accomplished prior to initial criticality of Unit 2.
Abnormal operating procedure 1-AP-47, " Unit Operation During Opposite Unit Emergency",
was approved May 28, 1980 to provide this guidance. Review of the procedure by the inspectors resulted in comments which were discussed with station management and subsequently incorporated into the procedure.
Station management and NRR personnel were informed that this commitment had been met.
7.
Procedure for Use of Public Address System Unit 2 operating License NPF-7 condition 2.D(5)v(7) requires that VEPCO establish procedures to limit the use of the public address system to plant related matters prior to Mode 2 operation. Station Administrative Procedure ADM 20.0, "Public Address System (GIA-TRONICS)" was approved May 28, 1980 to accomplish this function.
It establishes limits of usage for official business only and assigns priorities for use during emergency conditions.
Station management and NRR personnel were informed that the inspectors had no comments on the procedure and that this requirement of license condition 2.D(5)v(7) appeared to have been met.
8.
Followup on Other Previously Identified Open Items - Units 1 and 2 Differential Pressure Switches not Seismically Qualified (338/79-52-01 a.
and 339/79-58-06)
This matter invo'.ved the discovery of non-seismically qualified dif-ferential pressure switches being used in condenser cooling water controls for the units 1 and 2 contrci room air conditioning systems (Unit 1 LER 79 158 and a 10 CFR 50.55(e) report on Unit 2).
Qualified switches, Barton Model 288A, were installed per Unit I design change 79-S86 and Unit 2 E&DCR's P-2705-2, P-270A-2, FS -5306, and P-7178-2.
Review of the above documents in completed form along with associated test documents, and observaticn of the newly installed switches on April 22, 1980, revealed the new switches to be installed, calibrated and tested. These items are closed.
b.
Diesel Generator Fan Cooling System (338/79-38-01)
Paragraph 5.b (9) of these details discusses corrective actions on this problem for Unit 2.
The Unit 1 problem, reported by LER 79-110, was resolved in a similar manner. Two additional bolts were added to the fan drive shaft coupling to allow higier blade pitch settings per Unit 1 Design Change 79-S62.
Glycol was added via Unit 1 MR's, with the second diesel being completed on November 16, 1979. Also, the Unit I preventive maintenance program was revised to schedule instal-lation cf treated water in May and ethylene glycol in November of each year. Item 338/79-38-01 is close c.
Testing of Reactor Trip Breaker e atacts for Permissive P-4 (338/79-48-02 and 339/79-56-03)
As defined by Unit 1 LER 79-138 and Unit 2 10 CFR 21/50.55(e) report via letter no. 951 of November 19, 1979, the reactor trip breaker auxiliary contacts which feed permissive P-4 were not being routinely tested.
P-4 provides an interlock to enable or defeat the capability to manually reset and/or block safety injection (see FSAR Figure 7.2-2).
As discussed in IE Report 50-338/80-07, details paragraph 6.f, an interim procedure was used to test these contacts on Unit 1 prior to startup from the first refueling outage.
Periodic test procedures 1-PT-36.10 and 2-PT-36.10 were subsequently approved for use in performing the Westinghouse recommended surveillance on a monthly basis on both Units 1 and 2.
On February 22, 1980, Westinghouse submitted a revised test method to NRC via letter no. VPA-80-20. This revised method was incorporated into 1-PT-36.10 and 2-PT-36.10 on Ma rch 26, 1980. Review of these PT's by the inspector resulted in minor comments involving typographical errors. These comments were acknowledged by the cognizant supervisor and the procedures were subsequently revised.
On May 7, 1980, the inspector requested that completed copies of these PT's for each Unit be made available for his review. It was determined that 1-PT-36.10 had not been performed since January 1980 and that 2-PT-36.10 had not yet been performed. Station management was informed May 9, 1980, that failure to perform 1-PT-36.10, which has a frequency of 31 days appeared to be in noncompliance (Infraction) with Unit 1 Technical Specification 6.8.1 (item no. 338/80-19-02). Both PT's were perforined May 8,1980.
In that the reactor protection system surveil-lance had not been required for Unit 2 before May 8, the inspector had no further questions for that Unit. Items 338/79-48-02 and 339/79-56-03 are closed.
d.
Inoperable Containment Isolation Valves (338/78-40-02)
This problem was reported by Unit 1 LER 78-094 and involved the failure of two isolation valves to close upon demand. Maintenance reports (MR) were issued for investigation, which revealed problems) with the valve limit switches.
Design change 79-S19, initiated for another reason, modified the limit switch mounting brackets for all Unit 1.
valves of this type, thus correcting this prcblem also. Item 338/78-40-02 is closed.
Inoperable Isolation Valve (338/79-09-04)
e.
This valve failure, reported by Unit 1 LER 79-03, was repaired under MR N1-79-01022045. No generic problems were identified.
Item 338/
79-09-04 is closed.
.
-15-f.
Control Room Air Conditioning Compressor Switches (338/79-15-01 and 339/79-21-01)
This problem was last discussed in IE Report Nos. 50-338/80-11 and 50-339/80-09. As noted in paragraph 5.a of that report, seismically qualified switches had been installed, but incorrect reset setpoints had been used on the Unit I switches; no safety-related calibration procedures had been developed; and the switches had not been entered into the preventive maintenance program.
During the current inspection, the inspector reviewed Instrument Cali-bration Procedure (ICP) P-1-HV-2 (safety-related) which was revised Ma rch 12,1980, to provide for calibration instructions for the dif ferential pressure switches for both units.
It was also noted that station administrative procedure (ADM) 10.0 had been revised on the same date to incorporate the six switches to assure periodic calibration.
ICP-P-1-HV-2 was verified to specify the correct trip and reset set-points for all switches, but as of May 7, 1980, the Unit 1 ICP's had not been performed, thus, these were still in the range of 228 to 245 psig versus the correct range of 250-255 psig. Unit 2 item 339/79-21-01 is closed, but item 338/79-15-10 remains open. Management stated that recalibration of the Unit I switches would be accomplished during the next suitable shutdown.
g.
Return of Equipment to Normal by Special Test Procedures (338/79-20-07 and 339/79-28-12)
This item involved an inspector observation that administrative controls for special test and startup test procedures did not require that they return systems / equipment to a normal or otherwise desired operating status at the conclusion of testing. Subsequent review revealed that no change in controls was required for startup tests. The Nuclear Power Station Quality Assurance Manual Paragraph 5.5.8.17, was revised March 24, 1980 to provide the necessary instructions for special test procedures. Items 338/79-20-07 and 339/79-28-12 are closed.
h.
Use of Tags on Control Boards (338/79-20-08 and 339/79-28-13)
Instructions to operators in placement of tags on control boards was accomplished via Standing Order No. 29., issued May 19, 1979. Discus-sions with licensed operators revealed that they were familiar with the instructions.
Observations by the inspector revealed little potential for obscuring control board indications by use of tags.
Items 338/79-20-08 and 339/79-28-13 are closed.
i.
Procedures for Feeding Dry Steam Generators (Units 1 and 2)
As noted in IE Report 50-338/79-10, North Anna had no procedures for feeding a dry SG.
Operators interviewed felt that there would never be a need for doing this in that it is not probable that all three SGs
-16-would be boiled dry, but stated that it would be done if necessary.
The inspector was advised at that time that such a procedure was being developed.
During the current inspection, the inspector revieweo abnormal proce-dures 1-AP-22.1, 1-AP-22.3, 1-AP-22.5, 2-AP-22.1, 2.AP-22.3, and 2-AP-22.5 which had been revised November 29, 1979, to provide instruc-tions for bandling a dry SG. Also discussed is a situation where all three SGs are losing level concurrently. The inspector had no further que st. ions on this matter.
j.
Auxiliary Feedwater Pump Motor Fans (Unit 2)
This item was initially reported in accordance with 50.55(e) and is identified as open item 339/80-17-13.
After reviewing Deficiency Report 80-N, Nonconformance Report 2-008, Rework Control Forms M2-1251, E-14761, E-14760, Purchase Order 25825, and the results of special test procedure 2-ST-3 thru step 4.26, it appears that the repairs were made satisfactorily and these pumps perform as designed and therefore this item (80-17-13) may be considered closed. The fans on the pumps installed in Unit I were determined tc have been satisfactorily installed by successful completion of the 48-hour test on the pumps.
k.
Low Head Safety Injection Pump Bearing Wear (339/77-37-04)
This problem, reported per 10 CFR 50.55(e) and 10 CFR 21, was common to Unit I and 2.
Unit 1 operating license, NPF-4, included a condition 2.D. (3).1 which required confirmatory testing of the LHSI pumps to demonstrate long term operability following modifications to reduce bearing wear.
Results of these tests for Units 1 and 2 pumps were submitted to NRR for review. As noted in paragraph 6.3.4.2 of Supple-ment No. 10 to the Units 1 and 2 Safety Evaluation Report, these results were found acceptable.
During the current inspection, the inspector reviewed information associated with Purchase Order NA-93496R108, placed with the pump vendor, Ingersoll-Rand Company, which demonstarted that the bearing modifications described in VEPCO's report of November 3, 1977 (serial no. 554) had been completed. The inspector had no further questions and item 339/77-37-04 is considered closed.
1.
Nonconservatism in Analysis for Dropped Control Rod (338/79-15-07)
As discussed in paragraph 5.b.(1) administrative procedures have been implemented as corrective action for this problem. Therefore this item and LER 79-41 are considered close.
.
-17-Service Water Mode of Containment Cooling m.
This problem was reported by LER 79-141 for Unit 1 on November 28, 1979 with a followup report submitted on December 14, 1979.
VEPCO letter serial No. 179 dated December 28, 1979 stated that administra-tive controls would be implemented to prevent opening of these valves upon reset of an SI signal.
It was observed that no administrative controls were implemented.
Station management was informed May 16, 1980 that failure to implement edministrative controls for this mode of operations appeared to be a deviation from a written commitment (338/80-19-05). This comment was acknowledged and subsequently these valves have had their motor operator breakers opened and locked and the keys administrative 1y controlled.
9.
Initial Fuel Loading - Unit 2 The inspector observed initial fuel loading activities from the control room, fuel building and containment at various times during the period of this evaluation. The following areas were observed to determine conformance to license, administrative and procedural requirements:
Technical specifications requirements and license conditions a.
b.
Calibration of nuc' car instrumentation c.
Procedural prerequisites and initial conditions d.
Shif t manaing e.
Procedural compliance f.
Maintenance of Inverse Multiplication plots g.
Surveillance of monitoring instruments h.
Shift turnover i.
Access centrol to the fuel handling area j.
Shif t work schedules k.
Implementation and control of procedural changes 1.
Data taking The inspector had no questions in the areas observed during the evaluation.
Additional inspection efforts relative to initial fuel loading are documented in IE Report 50-339/80-2.
-18-10.
Precritical Test Witnessing - Unit 2 Portions of the following tests were witnessed by the inspectors during May 1980:
2-SU-9, " Full Length Contrcl Drive Mechanism Timing Test"
.
2-SU-11, " Rod Drop Time Measurement"
.
.
2-SU-13, " Rod Position Indication" 2-SU-14, " Pressurizer Spray and Heater Capability and Setting Continuo 1s
.
Spray Flow" (initial conditions only)
2-SU-15, " Reactor Coolant System Flow Measurement" (initial conditions
.
only)
2-SU-44, " Vibration and Loose Parts Monitoring Test"
.
2-0P-5.3, " Reactor Coolant System Leak Test"
.
ICP-FM-2-FM-1, "Incore Flux Mapping System"
.
Except as noted above, one or more of the following were observed for each test witnessed:
a.
Procedure of appropriate revision was available and in use by all cre; member's b.
Minimum crew requirements were met c.
All test prerequisities and initial conditions were met and/or those which are waived were reviewed / approved in accordance with requirements d.
Special test equipment required by the procedure was calibrated and inservice e.
Transient test data equipment required by the procedure was calibrated to a common time base f.
Test was performed as required by the procedure; changes to the proce-dure were made in accordance with procedure /TS requirements g.
Crew actions appeared to be correct and timely during the performance of the test. Adequate coordination existed h.
All data was collected for final analysis i.
Test results appeared to meet acceptance criteri.
-19-Additionally, the Shift Supervisors log, Control Room operators log, and chronological log for each test were reviewed periodically during the test period. Planning and scheduling meetings were attended on a daily basis as time permitted. Within the scope of the inspection, no items of noncompli-ance as deviations were observed.
11.
Startup Test Engineer Qualifications - Unit 2 The units 1 and 2 FSAR, section 14.0, prescribes qualifications required for startup test engineers.
In that VEPC0 contracted with Babcoch and Wilcox (B&W) to provide assistance for fuel loading and startup test program of Unit 2, the inspector reviewed the qualifications and training of five B&W individuals who were assigned as startup test engineers. Also, qualifi-cations of three employees of VEPCO's Nuclect Fuel Operatioas Group who will serve as startup test engineers were reviewed. Based on review of resumes and discussions with selected individuals, the inspector had no questions regarding qualifications.
VEPCO's NPSQAM, section 2, paragraph 5.3.2 requires that temporary service personnel be indoctrinated in VEPCO's quality assurance procedures as related to their specific tasks to be performed. An internal Technical Services Department memorandum of March 22, 1980 documented conduct of such training for several B&W employees and VEPC0 employees not previously designated as startup test engineers. On May 1, 1980, the B&W startup test engineer on the 4 p.m. to 12 midnight shift in charge of control rod drop testing (2-SU-11) was not on the March 22 list.
When station management was questioned regarding receipt of NPSQAM tiaining, the inspector was advised that the individual apparently had not received such training. No documentation of such could be produced. Management was informed May 9, 1980, that failure to conduct such training appeared to be a deviation (339/80-20-03) f rom VEPCO's " Topical Report - Quality Assurance Program -
Operations Phase" (VEP-3A), section 17.2.2.7 as implemented by paragraph 2.5.3.2 of the NPSQAM. The required training was subsequently conducted and documented in a memorandum ot May 16, 1980.
12.
Reactor Trip and Safety Injection - Unit 1 On May 22, 1980, at 2247 hours0.026 days <br />0.624 hours <br />0.00372 weeks <br />8.549835e-4 months <br /> Unit I tripped due to hi-hi level in steam generator
"C" due to malfunction of the feedwater regulating valve. On May 23 at 0145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br /> a safety injection (SI) was received from high steam flow and lo-lo Tavg. The high steam flow signal had beer received due to loss of a.c. vital bus 1-III. Lo-lo Tavg occurred after all three reactor coolant pumps (RCP) were shut down upon isolation of component cooling to their thermal barriers when bus 1-III was lost. A more detailed scenarios is presented in Unit 1 LER-80-047.
Of particular interest during this event was the loss of bus I-III, apparently due to a voltage spike on the "J" emergency bus ubich occurred when the reserve station service transformer (RSST) feeding "J" bus was disconnee'ed with the associated emergency diesel generator IJ concurrently supplying power to the
"J" bus and to the switchyard through the "A" RSST. Implacations of the loss of the vital bus will be reviewed in more detail (open item 338/80 '9-03).
'
.
-
.
-20-13.
Charging Pump Operation Following Secondary Side liigh Energy Line Rupture -
Units 1 and 2 On May 8,1980, Westinghouse informed NRC of a potential problem involving cer.srifugal charging purap under the provisions of 10 CFR 21. As noted in Unit 1 LER 80-42, following a secondary side high energy line rupture and subsequent safety injection initiation, the charging pump recirculation (miniflow) line isolation valves automatically close.
If the pressurizer power operated relief valves (PORV's) were not operable due to loss of offsite power, adverse environment inside containment, PORV in manual mode, or the PORV block valve in a closed position oue to PORV leakage, the RCS pressure would increase because of injection flow and core decay heat gener-ation until it reached the setpoint of the pressurizer code safety valses (2485 psig). At this high a back pressure, the charging pumps may not be able to supply suf ficient flow to prevent damaging the pumps. Plant-specfic calculations were being performed at the conclusion of the inspection. This will be reviewed in more detail (open items 338/80-19-04 and 339/80-20-04).