IR 05000338/1980011
| ML19309H818 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/25/1980 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19309H793 | List: |
| References | |
| 50-338-80-11, 50-339-80-09, 50-339-80-9, NUDOCS 8005190754 | |
| Download: ML19309H818 (10) | |
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UNITED STATES
NUCLEAR REGULATORY COMMISSION 80051907[
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o 101 MARIETTA ST., N.K, SulTE 3100 ATLANTA, GEORGIA 30303
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Report Nos. 50-338/80-11 and 50-339/80-09 Licensee: Virginia Electric and Power Company Richmond, Virginia 23261 Facility: North Anna Units I and 2 Docket Nos.
50-338 and 50-339 License dos. NPF-4 and CPPR-78 Insp:ction at North A a Site ear Mineral, Virginia Inspector:
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Dat'e Sigued Approved by:
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floF Chief, RONS Branch Date Sign'ed SUMMARY Inspection on February 4-19 and 25-29, 1980 Areas inspected Unit 1 Areas Inspected This routine inspection by the resident inspector involved 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on site in the areas of previously identified open items and reportable plant events.
Unit 1 Findings Within the two areas inspected, one apparent item of noncompliance was identified (Infraction-failure to perform a safety evaluation on a proposed test as required by Technical Specifications paragraph 8).
Unit 2 Areas Inspected This routine inspection by the resident inspector involved 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> on site in the areas of previously identified open items and items of noncompliance, reports per 10 CFR 50.55(e) and 10 CFR 21, and plant readiness for fuel loading.
Unit 2 Findings No items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted Licensee Employees R. A. Bergquist, Assistant Instrument Supervisor H. W. Burruss, Associate Engineer
- W. R. Cartwright, Station Manager
- J. D. Kellams, Superintendent - Station Operations J. R. Harper, Instrument Supervisor
- E. W. Harrell, Superintendent - Maintenance D. M. Hopper, Health Physics Supervisor R. T. Johnson, Construction QA Engineer
- W. R. Madison, NRC Coordinator
- E. R. Smith, Superintendent - Technical Services A. H. Stafford, Health Physicist Other licensee employees contacted included two technicians and three office personnel.
- Attended exit interviews.
2.
Exit Interview The inspection scope and findings were summarized on February 15 and 29, 1a80 for those persons indicated in Paragraph I above.
The apparent Infraction described in paragraph 8 was discussed in detail on February 29, 1980. The inspector's comments were acknowledged.
3.
Licensee Action on Previous Inspection Findings (closed) Infractions 339/79-28-09 and 339/79-35-09: Failure to Perform Timely Reviews of Potentially Reportable Items. Corrective measures defined in licensee responses of July 17, 1979 (letter no.
523) and September 10, 1979 (letter no.
667) were reviewed and discussed with station and corporate personnel. Various memoranda and meeting minutes documenting performance of reinstruction of personnel in reporting require-ments were also reviewed. These matters are considered closed.
Infraction 339/79-35-02 involved untimely review of the failure of Unit 2 motor control center 2-EP-MC-11, which was eventially reported per 10 CFR 50.55(e) and 10 CFR 21.
Inspection, repair and retest records, including the following documents, were reviewed:
Preoperational Deficiency Report (PDR) 2-106 a.
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Nonconformance and Disposition (N&D) reports 3792 and 3792A l
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Field QC Inspection Reports associated with the above N&Ds.
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Rework Control Form (RCF) E-14573, and A Klockner - Moller Corporation service report of May 31, 1979.
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The inspector had no further questions on this repair effort.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Followup of Previously Identified Open Items Control Room Air Conditioning Chillers (Items 338/79-15-01 and a.
339/79-21-01). Results of cfficiency testing of the station service water reservoir were submitted to NRC by letter no.
237D on March 8, 1979. These results noted that service water (SW) temperatures could be higher than previously assummed due to less efficient operation of the spray system in the reservoir. A problem associated with increased SW temperatures was potential overloading of the Units 1 and 2 control room chillers which use SW for condensor cooling.
Corrective action, implemented by design change (DC)19-515 for Unit I and Engineering and Design Coordination Report (E&DCR) P-2550-2 for Unit 2, involved addition of a pressure switch to unload some of the compressor cylinders upon high discharge pressure. This would assure continued operability of the chillers and maintain control room temperatures below maximum limits.
In addition to DC 79-515 and E&DCR P-2550-2, the following documents were reviewed and discussed:
(1) An S&W Trip Report for testing conducted on site January 22-26, 1979 (2) A Westinghouse Test Report dated February 2, 1979 (3) Instrument calibration procedire (ICP) P-1-MI-1, " Field Mounted Instrument Calibration Procedure," performed for the three Unit 1
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pressure switches on November 17, 1979.
(4)
ICP-P-1-MI-1 performed for the Unit 2 pressure switches on May 18-19, 1979 and (5) Purchase Order No. NA-1184-18.
Review of the above resulted in the obseivation that no reset setpoints for the Unit I switches had been specified in DC 79-515, although it was contained in E&DCR P-2550-2. Apparently because of this, the Unit 1 switches were set improperly, with a range from 228 psig to 245 psig versus the required 250 to 255 psig. Additionally, ICP-P-1-MI-I is a nonsafety-related procedure, whereas the switches serve a safety-related function. Also, at the time of review, the new pressure
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switches had not been entered into the station preventive maintenance program to assure periodic recalibration. During discussions with licensee personnel, the inspector was informed that the Unit 1 switches would be recalibrated, a new safety-related procedure would be developed, and the switches entered into the preventive maintenance program.
It was observed that the completed DC package had not yet received final
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review by all cognizant supervisors at which time determinations are made on the need for new procedures, revisions to procedures, etc., as defined by Section 3 of the Nuclear Power Station Quality Assurance
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Manual (NPSQAM).
Licensee personnel were informed that items 338/79-15-01 and 339/79-21-01 would remain open pending completion of the actions described above.
b.
AC and DC Solenoid Valves (Items 338/79-15-04 and 339/79-21-02).
As discussed in IE Report 339/79-21, a potential problem with AC and DC solenoid operated valves was identified on Unit 2 via a verbal 10 CFR 50.55(e) report on February 20, 1979. The problem involved a
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rework of these solenoids to replace their coils with ones with a
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higher temperature rating. During the rework effort on Unit 2, certain parts, such as retainer clips had been left out during reassembly, resulting in possible failure to operate. Unit I sole-i noids had apparently been replaced earlier with more complete quality control coverage, thus station management was confident that a similar problem did not exist in Unit 1, but preparations were being made for
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inspection of solenoids on both units to assure proper assembly and operation.
During the current inspection period, documents relating to inspection,
repair and testing of the solenoids were reviewed and discussed, including:
(1) Unit 2 E&DCR's 7167-2 and 6827C-2 i
(2) Unit 2 N&D's 3669 and 3688
(3) Unit 2 RCFs E-14377 and I-701 and associated inspection reports and checklists j
(4) Unit 1 maintenance report (MR) N1-79-04030845, and I
i (5) Unit 1 electrical maintenance procedure EMP-C-SOV-2, conducted
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April 7, 1979.
I Based on his review of the above documentation and discussions with
licensee personnel, it appeared that all Unit 2 solenoids had been inspected, repaired,'and retested as need;3, thus, item 339/79-21-02 is considered closed.
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The Unit 1 effort had not been completed and was continuing under control of MR N1-79-10111630, thus, item 338/79-15-04 remains open.
c.
Environmental Qualifications of Containment Isolation Valves (339/79-28-11)
As noted in IE Report 50-339/79-28, evaluation of environmental qualifi-cations of class IE equipment as required by IE Bulletin 79-01 had resulted in the discovery of a potential problem with six sampling system isolation valves. These are solenoid operated valves and the mechanism which could create a problen is a plastic plug which could block an air port in the solenoid actuator preventing proper operation. The solenoids in question were manufactured by ASCO (Model No. 8320A102). The valves are normally opened only when samples are required and are designed to fail in the closed position. This was reported as an LER on Unit I and per 10 CFR 21 and 50.55(e) on Unit 2.
Corrective actions were reviewed and an open item for that unit closed in IE report 50-338/80-07.
During the current inspection period, replacement of the Unit 2 valves with qualified ones was reviewed. E&DCR 6827C and RCF E-14377 controlled the replacement and retesting of the six (6) valves. Additionally, E&DCR 6827C recommended replacement of eight (8) ASCO solenoids for containment isolation valves on primary system sampling lines located in the auxiliary building in that they could be subjected to high post-accident radiation levels.
These were replaced along with the six in containment. New valves of the type having requisite qualifications were verified to have been installed by physical inspection. Licensee personnel were informed that there were no further questions on Unit 2 and that item 339/79-28-11 was closed.
Queationing related to the status of the eight similar valves for Unit 1 located in the auxiliary building revealed that they had not been replaced based on the position that they were not needed for post-accident sampling and that normal exposure would not be expected to exceed 100 Rads per ; ear (S&W 1etter to ASCO dated August 15, 1979). During discussions on this matter, it was noted that a qualified sampling system is to be installed as part of the long term effort of NUREG 0578 (see VEPCO letter No. 015 dated
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Janua ry 10, 1980). Acceptability of the eight solenoids for interim use will be evaluated further (open item 338/80-11-01).
d.
Protective Systems Process Sensing Lines (339/79-54-02)
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Unit 1 LER 79-142 discussed the discovery of two main steam pressure trans-mitters whose sensing lines were cross-connected. Corrective actions for Unit I were discussed in IE Report 50 433/d0-07. During the current inspection, investigative efforts f't esit 2 were reviewed. Protection system sensing l'.nes were verific4 o bn properly connected by procedure IMP-P-PROC-04, "*'3-? !ication ej l
'yi ec.ter Process Connections" and MR-N2-79 12110747. - All sensint line, attre found to be properly connected.
In addition to protection system instrumsats, the sensing lines for steam l
generator blowdown radiation monitors were chepked by IMP-C-PROC-23, a new version of the original procedure, and NR N2-80-01271930. This effort i
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found that all three sets of blowdown monitors / sensing lines were cross-connected. Deviation Reports 80-59 through 61 and PDR's 2-161 through 164 were initiated to correct these discrepancies. This had not been completed at the conclusion of the inspection period. Licensee personnel were informed that there were no further questions on Unit 2 protective system sensing lines (a fuel load item), but that open item 339/79-54-02 would be redefined to track the corrective action for steam generator blowdown monitors. This will require resolution prior to initial criticality.
6.
IE Bulletin 79-14, Unit 2 On February 7, 1980, licensee management reported that certain piping / pipe support pror'. ems had been identified during the review required by IEB 79-14, " Seismic Analyses for As-Built Safety-Related Piping Systems". This initial reporting was combined with problems on Unit 1 as LER's 80-01 and 80-16.
On February 21 and 27, additional Unit 2 problems were reported per 10 CFR 50.55(e) and per 10 CFR 21 for the latter case. These problems arose due to erroneous valve weights used in original stress calculations and involve calculated overstress conditions in piping supports / restraints during seismic events. Systems involved included reactor coolant, main steam, feedwater. safety injection, quench spray, chemical and volume control, and others. Open Item 339/80-09-01 is identified for followup purposes. Station management was informed that corrective actions must be completed prior to initial criticality.
7.
Unit 2 Status - Readiness for Operation Construction and preoperational testing of Unit 2 is essentially complete, although not all test results have been reviewed by station management.
IE Report 50-339/79-56, Details paragraph 9 listed activities to be completed and problems to be resolved prior to IE:II making a finding that Unit 2 would be ready for an operating license. A revised listing is given below, which reflects the addition of certain new problems recently reported and detection of other items closed in paragraph 5 of these Details. This listing does not include activities resulting from the Three Mile Island
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incident, such as those defined in NUREG 0585, "TMI-2 Lessons Learned Task Force Final Report", or those related to upgrading of emergency planning.
Items to be completed or otherwise resolved are categorized below by mile-stone:
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Items to be Resolved Prior to Fuel Loading a.
(1) Completion of preoperational testing and satisfactory resolution
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of test deficiencies. As of February 29, 1980 104 of 107 tests
and retests had been completed and the results of 100 aporoved by the Station Nuclear Safety Operating Committee.
(2) Complete development of operationul procedures recommended by R. G. 1.33.
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(3) Complete implementation of industrial security program in accor-dance with the North Anna Modified Amended Security Plan.
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(4) Resolve Quality Assurance program outstending items involving instrument calibration control (79-11-13 and 79-11-24).
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b.
Item to be Resolved Prior to Two Unit Operation in Modes 4 or Above.
Modify the electrical transfer buses to eliminate overloading during two unit operation (LER 79-57/0IT-0 for Unit 1; Unit 2 open item 79-28-10) 1/
Items to be Resolved prior to initial criticality c.
(1) Verify that no mechanical interferences exist between incore guide tubes or modified insulation supports and the reactor vessel lower head insulation due to thermal growth during precriti-cal testing (78-24-02).
(2) Visual verification of hydraulic snubber operability at operating temperature (79-26-02).
(3) Verify adequate seismic stress analysis of safety related piping per IE Eulletin 79-07 (79-28-07).
(4) Verify adequacy of service water and component cooling water pipe supports to resist thermal loadings over service water temperature range (78-36-01).
(5) Determine if atypical weld material was used in the reactor vessel seam welds (IE Bulletin 78-12A).
(6) Verify that piping and support loads are within limits for systems where erroneous Velan check valve weights were used (79-21-03; also see IE Bulletin 79-04).
(7) Resolve error found in DNBR analysis for single dropped control rod (79-28-01).
(8) Resolution of potential degradation of electrical penetrations due to overcurrent conditions (79-46-]1).
(9) Verify adequacy of emergency diesel generators fan cooling systems
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over all ambient temperature ranges (79-44-02).
(10) Replacement of cracked disc for containment isolation valve MOV-2380 (79-54-06).
1/ These items are complete and ready for inspection per licensee management.
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(11) Upgrade safeguards building exhaust ventilation system to Seismic Class I (79-54-06).
(12) Verification of adequacy of piping and piping supports for those systems wherein incorrect seismic response spectrum curves were used in stress analyses (79-56-02).
(13) Verification that reactor trip breaker auxiliary contacts for interlocks P-4 are adequately tested by solid state protection
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system surveillance procedures (79-56-03).
(14) Modification of circuitry to assure that engineered safety features equipment do not return to nornal operating modes following reset of containment depressurization actuation (CDA) or safety injection (SI) (79-55-01).
(15) Resolution of potential non-conservative feature of Westinghouse large break LOCA-ECCS evaluation model and its impact on peak clad temperatures (79-58-03).
(16) Resolution of non-conservatism involving error analysis of reactor negative rate trip for dropped control rods (79-58-04).
(17) Resolution of the non-conservatism in method used to generate power distributions and rod worths for dropped rod events (79-58-05).
(18) Verification of seismic quclification of control panels for the control room air conditioning chillers (80-05-01).
(19) Completion of modifications of piping supports / restraints to preclude overstress conditions during seismic events (IEB 79-14, item 80-09-01).
(20) Rerouting of steam generator blowdown sample monitor sensing lines found to be crossconnected (79-54-02).
d.
Items to be Resolved Prior to Exceeding 25% Power
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(1) Develop surveillance program for fiberglass spray pond piping and supports (R. G. 1.72).
(2) Complete environmental qualification testing and resolve any deficiencies identified for electrical components inside contain-ment (IE Circular 78-08/78-28-01 and IE Bulletin 79-01).
8.
Indications of Loose Parts - Unit I l
During preparations for startup from an unplanned reactor trip, indications i
of loose parts were received on the loose parts monitoring system (LPMS) on February 20, 1980.
Indications were received from the upper reactor vessel head and all three steam generators. Following investigative efforts and
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evaluations by Westinghouse and VEPCO, a conference call was conducted February 22, 1980 between NRC, Westinghouse, and VEECO. Safety evaluations which had been performed, and proposed investigations, surveillances and evaluations to be performed were discussed. Following return of Unit 1 to power, alarms were intermittent and short lived.
Investigations were continuing at the conclusion of the inspection period. This matter will be reviewed further and is designated as open item 338/80-11-02.
Following return to critical operations on February 22, 1980, noises were heard again in the steam generators and the unit was returned to Hot Standby for further evaluations. Believing that the main steam line non-return valves (NRV) were " perking" and creating the impacts observed, station personnel decided on February 23 to create conditions for " perking" one of the NRV's while monitoring the LPMS to ascertain if this was the source.
These valves, shown on FSAR Figure 10.3.2-2, are vertical plug types and
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will lift when sufficient steam pressure builds under the disk, then will rescat after pressure release, resulting in a " perking" motion. With two NRV's and their bypasses shut, one NRV was opened, its bypass shut, and a condensor steam dump valve opened to create an artifical pressure drop j
j across the NRV. According to station personnel, the valve perked for up to thirty (30) seconds when a safety injection (SI) was received due to differential pressure between main steal lines. The SI was reported as a prompt LER per Unit 1 Technical Specifications.
During discussions on the NRV test, it was learned that no evaluation of
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the proposed evolution, which is not in accordance with operating procedure 1-(0P)-28.1, " Main Steam System", had been performed by the Station Nuclear
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Safety and Operating Committee (SNSOC). Station management was informed that failure to have the proposed test reviewed by the SNSOC was contrary to Technical Specification 6.5.1.6b, which requires that the SNSOC review
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all tests and experiments which affect nuclear safety. The inspector also noted that 1-OP-28.1 was not followed, which is contrary to Specification 6.8.la in that a procedure deviation was not processed. This apparent i
Infraction (338/80-11-03) was discussed in detail during the exit interview l
on February 29, 1980.
9.
Contaminated Water Spill - Unit 1
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On February 26, 1980, a health physics technician observed water to be leaking from a pressure gauge sensing line on the discharge of a Unit I
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refueling water storage tank (RWST) chiller (FSAR Figure 6.2.2-1).
The leak was through a carbon steel fitting for the pressure gauge, which had apparently been installed on a temporary basis and not removed. Emergency Plan Implementing Procedure (EPIP-2) was implemented. Approximately 270 gallons of water (worst case) was assummed to have leaked onto the ground
by the RWST. Based on RWST activities, about 56 micro curies of radioactivity were released to the ground. The nearest stora dra'a was sampled and
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monitored with negative results, indicating that no offsite release, or release to an unrestricted area was made. The event will be reported as a thirty (30) day Licensee Event Report per Technical Specifications. Circum-
stances surrounding installation of the carbon steel fitting had not been A
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determined at the conclusion of the inspection. This will be reviewed in more detail (open item 338/80-11-04).
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