IR 05000338/1978045
| ML19305A312 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/23/1979 |
| From: | Cantrell F, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19305A309 | List: |
| References | |
| 50-338-78-45, 50-339-78-38, NUDOCS 7903130200 | |
| Download: ML19305A312 (1) | |
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UNITED STATES
- p ># R80 NUCLEAR REGULATORY COMMISSION o,,
REGION il
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c 101 M ARIETTA sTRE ET. N.W.
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ATLANTA, GEORGI A 30303 s,...../
Report Nos.:
50-338/78-45 and 50-339/78-38 Docket Nos.:
50-338 and 50-339 License Nos.: NPF-4 and CPPR-78 Licensee: '/irginia Electric and Power Company Post Office Box 26666 Richmond, Virginia 23261 Facility Name: North Anna Pcwer Station, Units 1 and 2 Inspection at: North Anna Power Station, Mineral, Virginia Inspection conducted:.Dec[mber 11-2, 1978
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Inspector:
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M. S. Kidd, ResiMent Inspector Date Signed Accompanying Personnel: Ncne Approved by:
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F. STCantrell, Act ig nief Date Signed Reactor Projects Sec on No. 2 Reactor Operations and Nuclear Support Branch Inspection Summary Inspection on December 11-29, 1978 (Report Nos. 50-338/78-45 and 50-339/78-38)
Unit 1 Areas Inspected:
Routine inspection by the Resident Inspector of licensee event reports, follow-up on IE Circulars, environmental reporting, and organizational changes.
The inspection involved nine man-hours on-site by the NRC Resident Inspector.
Unit 2 Areas Inspected: Routine inspection by the Resident Inspector of test engineer qualifications, fuel storage, plant tours, IE Circulars, preoperational testing, plant status, organizational changes, and follow-up on a previously identified, unresolved item. The inspection involved 25 man-hours on-site by I
the NRC Resident Inspector.
Results: Of the ten areas inspected, no items of noncompliance or deviations were identified.
7903130.200
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RII Rpt. Nos. 50-338/78-45
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and 50-339/78-38
. -1-DETAILS 1.
Persons Contacted Licensee Employees H. W. Burruss, Assistant Engineer
- W. R. Cartwright, Station Manager W. F. Diehl, QC Engineer I. B. Ferrer, Assistant Engineer P. A. Furman, Assistant Engineer G. A. Karm, Engineer
- J. D. Kellams, Superintendent-Station Operations M. A. Harrison, QC Engineer
- S. L. Harvey, Acting Operations Supervisor D. C. McLain, Engineering Supervisor P. A. Slatter, Resident QC Engineer-Construction
- D. L. Smith, Resident QC Engineer-Operations
- E. R. Smith, Supervisor-Engineering Services R. C. Starr, Shift Supervisor F. C. White, Acting Fire Marshall H. K. Wong, Assistant QC Engineer
- D. C. Woods, NRC Coordinator Other Organization (Stone and Webster)
G. M. Brynes, Senior FQC Engineer-Audits and Coordinators Group C. S. Majumdar, Senior FQC Engineer-Electrical S. J. Patton, Senior Engineering Clerk-Release Status Facility E. W..c-uell, Electrical Engineer-Site Engineering Office
- Attended management interview sn December 15, 1978
- Attended management interview on December 15 and December 22, 1978 2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (339/77-37-01):
Review of Unit 2 Engineering and Design Coordination Reports. See paragraph 5 of these Details.
3.
Unresolved Items i
Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve noncompliance or deviations. Na new unresolved items were identified during this inspectio.
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-2-4.
Management Interviews The inspection scope and findings were summarized on December 15 and 22, 1978 for those persons indicated in paragraph 1.
5.
Review of Unit 2 E&DCR's This unresolved item (339/77-37-01) was initially discussed in IE Report 50-339/77-37, Details I, paragraph 6.g(3).
As noted in that report, VEPC0 was to review all Engineering and Design Coordination Reports (E&DCR) written on safety-related (Category I) equipment to verify that they had been completed or otherwise reco1ved. During thi: inspection, the review programs of VEPCO and S&W were reviewed and certain E&DCR's were randomly selected to confirm closure or system for tracking and to confirm proper categorization. Findings of another inspector involved in this effort are given in IE Report 50-339/78-36, Details I, paragraph 2.
a.
Program Review Findings relative to review of the VEPC0/S&W program are given in Report 50-339/78-36.
b.
Independent Review to Assure Completion or Resolution The inspector randomly selected Category I E&DCR's from the Unit 2 master list for review of documentation to verify that those which were indicated closed had been closed, and those which were indicated open were in active files for processing.
Those selected are enumerated below, with subject matter covered.
P-526 Clarification of Specification NAS-347
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P-1922M Clarification of E&DCR P-19226
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PS670-2 Piping-Minimum Wall Thickness Calculationc
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6296-1 Heat Tracing Thermocouples
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10849 Aluminum Cable Termination
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11955 Testing of Cables at Penetrations
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I pnaa Weld Procedure W-37-B
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J419s Hydrostatic Testing
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RII Rpt. Nos. 50-338/7e-45 and 50-339/78-38-3-22153-1 Slip-On Orifice Flanges
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23046-1 Hanger Standards
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23268-1 Piping Classification
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23443-1 Cleaning of Piping Prior to Insulation
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24781-1 Revision to PQM Manual
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25147-1 Coating of Stainless Steel Surfaces
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30911 Painting for Corossion Protection (FSAR
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Comment 6.4)
31531A Spacers under Base Plates
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32011-2 Restraint for Line SI-538
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31914-2 Hangers for Quench Spray Lines
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PS-0560 Revision of Specification NAS 49-1049
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14703 FSAR Versus Specification NIS 341
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Review of the above dacumentation revealed that all but the last two were closed with appropriate signatures.
PS-0560 was still open with work pending. Number 14703 involved a revision to FSAR Section 8.3.1.1.2.4.
Review of the FSAR revealed that it had been revised via Amendment 61 on July 7,1977. An S&W representative was asked why the E&DCR would still be open even though the work requested had been completed. In that the work had been completed and the E&DCR was in an active file, the inspector had no further questions at that time. He was informed on December 20 that the E&DCR had been closed.
c.
Review for Proper Categorization The Category II and III E&DCR's listed below were randomly selected for review to verify that the proper safety category had been assigned.
4919-1 Revision to Specification NAS-11
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-4-05971 Waterproofing containment
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11678 Use of Crimping Tool
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12574A Rewiring Pressure Switches
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14358A Instrument Identification
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15344 Specification NAS 462
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20794 Specifications NAS 244 and 264
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21424 Specification NAS 414
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22017 Material Specification-Hangers
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24004 Insulation
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24773 Specification Clarification
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P-1546 Circulating Water System
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During the review, safety classifications given in Section 2 of the Nuclear Power Station Quality Assurance Manual, S&W's 431 equipment sort. and the inspector's knowledge of plant systems were used as a basis for determining appropriate categorizations. No discrepancies were noted.
Licensee personnel were informed that this unresolved item (339/77-37-01) was considered closed.
6.
Qualification of Test Engineers-Unit 2 Education and work experience requirements for Unit 2 preoperational test engineers are given in Section 14.0 of the Units 1 and 2 FSAR.
These requirements are implemented by station administrative procedure (ADM) 102.0, " Qualifications lof Pre-operational Testing Personnel",
which also defines on-site training requirements for the test engineers.
Records of previous education and work experience, along with on-site training records, were reviewed December 20, 1978 for six engineers prcviously assigned to conduct Unit 2 pre-operational tests, who had not
participated in the Unit 1 test program. These records were compared to the requirements of ADM 102 and FSAR Section 14.0 with no discrepancies resulting. Qualifications of test personnel involved in the Unit 1 test program were not verifie _
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-5-One of the six engineers was interviewed to confirm receipt of the on-site training.
Discussions with him confirmed receipt of the training and demonstrated an understanding on his part of the administrative controls for conduct of the test program. The inspector had no comments in this area.
During the review of educational and training records, the inspector observed that three engineers hired since December 27, 1977 have not yet been certified per ADM 102 due to lack of one year's experience. They will be certified after the one year's experience has been obtained, according to their supervisor.
7.
Unit 2 Fuel Storage Approximately two-thirds of the fuel needed for initial fuel loading of Unit 2 is presently stored in the new fuel racks in the Unit I and 2 fuel building (see FSAR Figures 1.2-17 and 1.2-18).
On December 19, 1978, the inspector visited the fuel building to verify that provisions for physical security and environmental protection of the fuel were in agreement with those of the application for the Unit 2 Special Nuclear Materials (SNM) License dated October 3,1977 (serial no. 148).
Security measures were found to be equal to or greater than the commitments of the SNM application and the Station Security Plan.
Protection of the fuel from dust, debris and physical damage was also found to be adequate. Observation of fire extinguishers in the area revealed them to have recent inspection dates. On extinguisher on the north wall of the building had been splattered with silicone foam used to seal penetrations. Inspection by the Fire Marshall revealed it to be operational.
Housekeeping was in order except in the immediate vicinity of rework on the security systems for door alarms and locks. Efforts to maintain housekeeping in that vicinity appeared reasonable.
Within the areas inspected, no items of noncompliance or deviations were identified.
8.
Plant Tours Tours of selected plant areas were conducted December 19 and 21, 1978, with emphasis on Unit 2.
During the tours, the following items, as I
available, were observed:
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-6-a.
Hot Work Adequacy of fire prevention / protection measures used.
b.
Fire Equipment Operability and evidence of periodic inspection of fire suppression equipment.
c.
Housekeeping Minimal accumulations of debris and maintenance of required clean-liness levels in systems under or following testing.
d.
Equipment Preservation Maintenance of special preservative measures for installed equipment as applicable, e.
Component Tagging Implementation and observance of equipment tagging for safety or equipment protection.
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Instrumentation Adequate protection for installed instrumentation, g.
Cable Pulling Adequate measures taken to protect cable from damage while being pulled.
h.
Communication Effectiveness of public address system in all areas toured.
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Equipment Controls Effectiveness of jurisdictional controls in precluding unauthorized work on systems turned over for initial operations or pre-operational testing.
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Foreign Material Exclusion Maintenance of controls to assure systems which have been cleaned and flushed are not re-opened to admit foreign material.
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RII Rpt. Nos. 50-338/78-45
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Security Implementation of security provisions.
Particular attention to maintenance of the Unit 1/ Unit 2 interface and that for the fuel building.
1.
Logbooks The chronological logs for Unit 2 pre-operational tests 2-PO-39.3,
" Residual Heat Removal System Valve Interlocks Test", 2-P0-37.1,
" Quench Spray System Pump Performance Testing", and, 2-P0-1,
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" Emergency Power System", were reviewed as the respective tests were in progress on December 21 and 22,1978.
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Testing Portions of the testing conducted per 2-P0-1, 2-P0-37.1, and 2-PO-39.3 were observed on December 21 and 22,1978. Findings are discussed in paragraph 9 of these Details.
Except as discussed below, no discrepancies were noted for those activities observed when compared to the applicable NAS standard, Unit 1 Technical Specifications, or other guidance, such as the station security and emergency plans.
During the tour on December 21, the inspector observed that the door from the Unit 2 cable vault to the emergency switchgear and relay room at elevation 254 feet in the service building (see FSAR Figure 1.2-22) had been tied open. Although workers were in the switchgear room, there was no firewatch evident. This door, along with an identical one on the Unit I side of the switchgear rooms, serve as fire barriers and as part of the control room pressure envelope boundary for emergency ventilation purposes.
As fire barriers, these doors are subject to Unit 1 Technical Specification 3.7.15, which states that all " penetration" fire barriers protecting safety-related areas shall be functional at all times and requires that s fire watch be established within one (1) hour if a barrier becomes non-functional.
The Unit 2 door was closed by station personnel immediately after being notified of its condition by the inspector.
The length of time it had been open could not be established by the inspector.
I These doors, as well as others in the plant which. isolate areas protected by carbon dioxide fire suppression systems, can be lef t open by placing a clip, chained to the door, over the top of a carbon dioxide actuated plunger. Upon receipt of a fire alarm and automatic gas release, the plunger forces the slip off, allowing
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-8-the door to close.
In the case of these two doors, they could be left open on the carbon dioxide plunger, but would close automati-cally only upon receipt of a fire alarm in the cable vault (switchgear/ control room complex is not carbon dioxide protected).
This design would allow the doors to stand open under certain plant conditions which require implementation of the control room complex emergency ventilation (bottled air) system, which is designed to maintain a positive pressure in the complex for one (1) hour as discussed in FSAR Section 9.4.1.
Following discussions on the problem, station personnel removed the chains and clips from the two doors plus another one leading the cable spreading room over the control room on December 22, 1978.
In addition to removing the door chains, operator log 1-LOG-GC will be used to document inspections of the doors every four hours to assure they are closed. Following upgrading of the door latching mechanisms to comply with new security requirements, they will be accessible only by use of card readers and will be alarmed thus providing better control.
In the interim, station management has initiated actions to assure that all construction workers are re-instructed on the significance of these doors, and the need to keep them closed. The inspector had no further questions on the subject at that time.
9.
IE Circulars The following IE Circulars were reviewed to verify that they had been received by station management, reviews for applicability to Units 1 and 2 had been perfonned, and appropriate corrective action had been taken or planned.
a.
IE Circular 78-05, Inadvertent Safety Injection During Cooldown This IEC was issued following an SI on a four loop Westinghouse plant.
It was initially discussed by the Station Nuclear Safety and Operating Committee (SNSOC) on June 8, 1978. On December 18, the SNSOC reviewed and concurred with a Westinghouse letter on the subject, dated December 5, 1978, which concludes that a three-loop plant such as North Anna 1 or 2 is not subject to the same uneven loop flows as four-loop plants, and that the common residual heat release headers used in North Anna 1 and 2 provide further assurance
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of maintaining similar steam header pressures. The inspector had no further questions on this IE.
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-9-b.
IEC 78-18, UL Fire Test This IEC was issued for review only.
The inspector confirmed review of the IEC by station personnel through review of SNSOC minutes for meeting 78-179 on November 27, 1978. There were no questions on this matter.
10.
Environmental Reporting-Unit 1 Due to confusion at certain other facilities concerning information required to be repcrted per Appendix B Technical Specifications, this matter was discussed with station personnel to assure that their understanding was correct.
The discussions revealed that Unit 1 reporting requirements are well understood.
11.
Test Witnessing-Unit 2 a.
Scope Portions of three pre-operational tests were witnessed as discussed below. The following observations were made for each of the tests.
(1) Latest revisions of test procedures were available and in use.
(2) Test prerequisites appeared to be met based upon observations of selected ones for each test.
(3) Portions of tests observed were performed per procedures.
(4) Chronological logs were used to document significant events, discrepancies, and interruptions.
(5) Crew actions appeared to be correct.
(6) Data was collected as required.
(7) Procedure reviews and approvals were in accordance with the Nuclear Power Station Quality Assurance Manual (NPSQAM).
(8) Drawings in use were current issues.
(9) Test engineers had received training required by station
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administrative procedures (see paragraph 6).
(10) Special test equipment was calibrated and in service, as require..
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-10-(11) A member of the QA staff was present and observing the conduct of each test.
b.
2-P0-1, Emergency Power System Portions of the testing per this procedure on the emergency diesel generators was observed December 22, 1978. Section 4.8, which was to verify the capacity of one air receiver (2JA) to start the diesel generator five (5) times with the compressor shut down was in progress. On the second and third starts, conducted manually from the local panel per operating procedure 2-0P-6.2, the engine tripped due to overspeed. The test was interrupted to investigate this problem and was re-run at a later date (inspector not on-site).
During the starts on December 22, the air receiver pressure dropped from 202 psig to 115 psig; thus, it appeared that two more starts would have been feasible. Results of this test as well as other significant ones will be reviewed by IE inspectors as part of the routine inspection program.
c.
2-P0-37.1, Quench Spray System Pump Performance Testing (Re-Run)
Steps 4.1.1 and 4.1.2 of 2-P0-37.1 were re-run December 22, 1978 to re-verify pump flows on recirculation to the refueling water storage tank. The re-run was necessary due to an error in valve lineup in the original test. Data obtained was used to verify pump performance and establish baseline data for in-service testing. Final calculations and data reduction will be reviewed as part of the routine inspection program.
d.
2-P0-39.3, Residual Heat Removal System-Valve Interlocks Test This test was attempted on December 21, 1978, but was interrupted when the pressure comparator controlling valve MOV-2700 did not re-set at the proper setraint. The test was repeated at a later date (inspector not on-site). Results of the test will be reviewed as part of the routine inspection program.
12.
Plant Status-Unit 2 On Decer' er 19, 1978, the inspector was informed that VEPCO's offical estimate for fuel loading of Unit 2 was April 15, 1979. Unit 2 systems, I
subsyste s and buildings have been divided into about 227 groups for release t ) VEPCO operations for pre-operational testing and fuel loading.
As of Derember 29, 1978, some 110 of the 227 " conditional" releases had been ef fe. te t...--
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-11-A total of 93 pre-operational tests have been identified.
As of December 29, all but seven of these had been written and approved.
Approximately twelve tests were in progress or had been completed. Hot functional testing was scheduled for mid to late January.
13.
Station Reorganization On December 20, 1978, the inspector was informed by the station manager that the station organization shown in Unit 1 Technical Specification Figure 6.2-2 would be changed effective January 1,1979, if no objections were made by NRC. The change in organization, depicted in a letter of December 26 to NRR and IE (serial number 652), involved the creation of three superintendent positions for operations, maintenance and technical services, and a new supervisor of administrative services. The changes are expected to streamline station operations while retaining all required functions and responsibilities. Licensee representatives were informed that IE had no concerns regarding the change, but that formal approval would come from NRR.
Other changes were also proposed in the referenced letter, including composition of the Station Nuclear Safety and Operating Committee (SNSOC).
The inspector was assured by the station manager on January 2, 1979, that the SNSOC composition would not be changed until the Technical Specification change is approved.
14.
Licensee Event Report Review (Unit 1)
The following LER's 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.
a.
LER 78-099/03L-0, Steam Generator High Level Trip Setpoint Nonconse rvative b.
LER 78-101/03L-0, Power Range Detector N-44, Placed in Trip Following Detector Calibration until Recalibrated LER 78-102/03L-0, Steam Generator Level Transmitter Inoperable c.
d.
LER 78-104/03L-0, Auxiliary Feedwater Pump Out of Service I
LER 78-105/03-L-0, Axial Flux Difference Out of Target Bands e.
f.
LER 78-106/03-L-0, Faulty Gain Pat on Power Range Channel b_
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RII Rpt. Nos. 50-338/78-45 and 50-339/78-38-12-The inspector had no comments on the LER's except for 78-104. The cause code for this was
"X" (other), while the inspector felt that use of an
"E" (component failure) was more appropriate in that steam leakage from a valve resulted in inoperability of the pump for repairs to the valve.
During discussions on this matter with station personnel, the inspector was informed that a corrected report would be submitted (open item 338/78-45-01).
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