IR 05000338/1979048
| ML19344D087 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/28/1979 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19344D080 | List: |
| References | |
| 50-338-79-48, 50-339-79-56, NUDOCS 8003110207 | |
| Download: ML19344D087 (10) | |
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UNITED STATES a*^
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NUCLEAR REGULATORY COMMISSION n
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REGION 11
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101 MARIETTA ST N.W SUITE 3100 e
ATLANTA, GEORGIA 30303
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Report Nos. 50-338/79-48 and 50-339/79-56 Licensee: Virginia Electric and Power Company Richmond, Virginia 23261
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Facility Name: North Anna Units I and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and CPPR-78 Inspection at: North Anna Site near Mineral, Virginia
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Inspector:
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M. S. Kidd Resident Inspector Date Signed Approved by:
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o P. J. Kellogg, Section hief, RONS Branch Dat6 Si'gned'
SUMMARY Inspection on November 5-13 and 26-30,1979 Unit 1 Areas Inspected This routine inspection by the resident inspector involved eight hours on site in the areas of IE Circulars and Notices and Licensee Event Reports.
Unit 1 Findings:
No items of noncompliance or deviations were identified.
Unit 2 Areas Inspected This routine inspection by the resident inspector involved fifteen hours on site in the areas oi IE Circulars and Notices, licensee reports per 10 CFR 50.55(e)
and 10 CFR 21, and unit 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
- W. R. Cartwright, Station Manager
- L. O. Goodrich, Mechanical Maintenance Supervisors G. A. Karn, Engineering Supervisor
- J. D. Kellams, Superintendent of Station Operations M. A. Harrison, Construction QC Engineer
- W. R. Madison, NRC Coordinator
- E. R. Smith, Superintendent of Engineering Services
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on November 9 and 30, 1979, for those persons indicated in Paragraph 1 above. Licensee personnel in attendance on November 30 were questioned as to whether the Licensee Event Reports for Unit I discussed in Paragraph 8.d through 8.f had been reviewed for applicability to Unit 2.
It was not known at that time whether such reviews had been conducted by Power Station Engineering.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Plant Status The first Unit I refueling fuel shuffle was completed October 30, 1979.
The shutdown is expected to continue through mid to late December, according to station management. Various design changes, modifications, and repairs are ongoing. Unit 2 preoperational testing was essentially complete at the conclusion of the inspection period, but not all tests had received final station review and approval.
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IE Circulars and Notices
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The following IE Circulars (IECs) and Notices (IENs) were reviewed to verify that they had been received by station management, if appropriate; t
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-2-reviews for applicability to Units 1 and 2 had been performed; and appro-priate corrective action taken as planned if needed:
a.
The following Circulars and Notices are closed based on documentary evidence reviewed indicating review and evaluation by station manage-ment and performance of actions required on other bases defined below:
IEC 79-04, Loose Locking Nut on Limitorque Valve Operators.
Securing of locking nuts on Unit I valves was documented by maintenance request (MR) N1-79-04090908. Unit 2 valves had not yet been checked. This will be inspected at a later date (open item 339/79-56-01).
IEC 79-08, Attempted Extortion - Low Enriched Uranium IEC 79-13, Replacement of Diesel Fire Pump Starting Contactors.
Licensee personnel determined that Units 1 and 2 do not use the types of coils and switches described in the IEC.
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IEC 79-17, Contact Problem in SB-12 Switches on General Electric Company Metalclad Circuit Breakers. Licensee personnel verified that none of these breakers are used in Units 1 ard 2.
- IEC 79-18, Proper Installation of Target Rock St.iety-Relief Valves. This type of valve is not used in Units 1 and 2.
- IEC 79-20, Failure of GTE Sylvania Relay Type PM Bulletin 7305 With 120 Volt AC Coil. These are not used in Units 1 and 2.
- IEN 79-01, Bergen-Paterson Hydraulic Shock and Sway Arrestor.
These are not used in Units 1 or 2.
- IEN 79-04, Degradation of Engineered Safety Features.
This Notice is administratively closed in that it was superceded by a letter from Nuclear Reactor Regulation (NRR) to the licensee on
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August 8, 1979.
- IEN 79-06, Stress Analysis of Safety-Related Piping. This Notice is administratively closed in that it was superceded by IE Bulletin 79-07. Licensee actions will be inspected during followup of the Bulletin.
- IEN 79-13, Indication of Low Water Level in The Oyster' Creek
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IEN 79-18, Skylab Reentry
IEN 79-19, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants. This Notice was superceded by IEB 79-17.
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IEN 79-21, Transportation and Commercial Burial of Radioactive Material
IEN 79-22, Qualification of Control Systems. This Notice was superceded by an NRR letter of September 17, 1979. This is also tracked as Unit 1 LER 79-120 and a similar report per 10 CFR 50.55(e) and 10 CFR 21 for Unit 2.
- IEN 79-23, Emergency Diesel Generator Lube Oil Coolers. Station Engineering Study 79-63 determined no problem exists for the Units 1 and 2 coolers.
b.
The following IECs and IENs are closed in that they are not applicable to North Anna 1 and 2 and were not received by the licensee:
IEC 79-01, Administration of Unauthorized Byproduct Material to Humans
IEC 79-06, Failure to Use Syringe and. Bottle Shields in Nuclear Medicine l
IEC 79-07, Unexpected Speed Increase of Reactor ~ Recirculation MG Set Resulted in Reactor Power Increase
IEC 79-14, Unauthorized Procurement and Distribution of Xenon 133
IEC 79-16, Excessive Radiation Exposures to Members of The General Public and a Radiographer
IEN 79-02, Attempted Extortion - Low Enriched Uranium
IEN 79-16, Nuclear Incident at Three Mile Island These Circulars and Notices were still under review by station personnel c.
at the conclusion of the inspection, and will be inspected at a later date:
IEC 78-09, Arcing of General Electric Company NEMA Size 2 Con-tactors
IEC 79-05, Moisture Leakage in Stranded Wire Conductors
IEN 79-03, Limitorque Valve Geared Limit Switch Lubricant
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IEN 79-10, Nonconforming Pipe Support Struts
IEN 79-11, Lower Reactor Vessel Head Insulation Support Problem No items of noncompliance or deviations were identified.
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Plant Tours Tours of selected plant areas were conducted on November 5, 8, 27 and 30, 1979. The following items, as available were observed:
a.
Fire Equipment Operability and evidance of periodic inspection of fire suppression equipment.
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b.
Housekeeping Minimal accumulations of debris and maintenance of required cleanliness levels in systems under or following testing.
c.
Equipment Preservation Maintenance of special preservative measures for installed equipment as applicable.
d.
Component Tagging Implementation and observance of equipment tagging for safety or equipment protection.
e.
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 preoperational testing.
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Maintenance of controls to assure systems which have been cleaned and flushed are not reopened to admit foreign caterial.
h.
Security Implementation of security provisions for both units.
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Testing On November 30, the inspector observed the lineup for testing of the Unit 1 motor driven auxiliary feedwater pumps in accordance with VEPCO letter No. 825 to NRR dated November 2, 1979. This endurance test was
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being conducted by special test Procedure 1-ST-13.
It was observed
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that a copy of the procedure was in use and that required data were being taken. There were no questions by the inspector.
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Protection of Installed Instrumentation:
Ori November 27, the inspector observed that the flexible conduit to steam generator level transmitter j
LT-2485 (Unit 2 containment) was broken at the transmitter junction point. Repairs were completed under the station maintenance program on November 30, 1979. There were no further questions.
Within the above areas, no items of noncompliance or deviations were observed when compared to the applicable station programs and procedures.
8.
Licensee Event Reports - Units 1 and/or 2 The following events or problems were reported as prompt (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) LER's per Unit 1 Technical Specifications and/or under the provisions of 10 CFR 50.55(e) and 10 CFR 21 for Unit 2 during the inspection period:
O a.
Seismic Analysis Errors (LER 79-151)
On November 5, 1979, VEPCO reported that Stone and Webster (S&W) had discovered that an incorrect amplified response spectrum curve had been used in pipe stress analyses for certain Unit I and 2 systems and that the wrong input format had been used for other seismic curves.
This matter was inspected at S&W corporate offices the week of November 24, 1979 and results reported in Vendor Inspector Report 999-00509/79-05. Recalculations were in progress at the inclusion of the inspection period. Open items 338/79-48-01 and 339/79-56-02 are identified for followup of corrective measures required.
b.
Testing of Reactor Trip Breaker Contacts for Permissive P-4 (LER 79-138)
On November 9, 1979, VEPCO reported that Westinghouse had informed them that the reactor trip breakers auxiliary contacts which feed permissive P-4 are not tested by routine surveillance procedures. P-4 provides an interlock to enable or defeat the capability to manually reset and/or block safety injection. This problem was reported for Unit 2 per 10 CFR 50.55(e) and 10 CFR 21 on November 14. Open items 338/79-48-02 and 339/79-56-03 are identified for followup of this problem.
Safety Function Reset Design Errors (LER 79-14)
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On November 7, 2979, VEPCO reported that errors had been discovered in the reset logics for containment depressurization actuation (CDA) and safety injection (SI) signals in that when they are reset, certain engineered safety features equipment would return to non-accident modes of operation without further operator action. This is not a
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6-I consistent with FSAR Supplementary Section 7.4.
Additional examples were reported later as they were discovered. This problem was reported for Unit 2 on November 16 under the provisions of 10 CFR 50.55(e) and 10 CFR 21. This problem was also reviewed with S&W during the inspection
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referenced in paragraph 8.a.
Items 338/79-46-01 and 339/79-55-01 are l
identified for followup purposes.
d.
Fuel Rod Burst Calculations (LER 79-143)
VEPCO reported on November 9, 1979, that Westinghouse had identified a potentially non-conservative feature in its large break LOCA-ECCS evaluati.on model involving full clad heatup rates. This could affect the FQ limit. Open item 338/79-48-04 is identified for a more detailed review of this problem on Unit 1.
As of November 30, this item had not been reported on Unit 2.
Applicability to Unit 2 will be reviewed during the next inspection period.
Nonconservatism in Error Analysis For N. gative Rate Trip (LER 79-152)
e.
On November 16, 1979, VEPCO reported that Westinghouse had discovered a non-conservative feature in the error allowances assumed in the analysis for the power range neutron flux high negative rate trip. It was assumed that differentiation of the NIS power signal would eliminate all steady state errors in the circuit before the differentiaton; therefore, only precision and accuracy errors after the differentiation need be considered. This assumption is non-conservative since errors associated with precision (hysterisis, repeatability, etc.) may not be entirely eliminated by differentiation. When these additional errors factored into the analysis, a reactor trip may not occur in some are cases. One method of preventing power overshoot if a
?. rip did not occur would be to limit control rod insertion when rods are in auto-
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matic above ninety percent power.
This will be reviewed in more detail (open item 338/79-48-05).
This ' problem had not been reported for Unit 2 as of November 30, 1979.
f.
Improper Analysis of Reactivity Profiles During Dropped Rod Events (LER 79-153)
On November 16, 1979, VEPCO reported that Westinghouse had determined that a more conservative method of generating power distributions and
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rod worths for dropped rod events should be used to preclude a possible power overshoot and DNBR limit violation. Interim corrective measures are expected to be the same as for item 8.e above. This will be reviewed in more detail (open item 338/79-38-06). As in the case of
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the two previous LER's for Unit 2 as of November 30, 1979.
9.
Unit 2 Status - Readiness for Operation Construction of Un'it 2 is essentially complete and preoperational testing has been completed, although not all test results have been reviewed by
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s station management. IE Report 50-339/79-43, Details, paragraph 11 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. Resolution of certain old items was addressed in IE Report nos. 50-339/79-41, 44, 46 and 54.
This listing does not include activities resulting from the Three Mile Island incident, such as those defined in NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report",
or those relating to upgrading of emergency planning. NRR has lead respon-sibility for followup of those items.
Items remaining to be completed or otherwise resolved are categorized below by milestone:
Items to be Resolved Prior to Fuel Loading a.
1)
Completion of preoperational testing and satisfactory resolution of test deficiencies. As of November 30, 1979, 99 of 103 tests and retests had been completed and the results of 84 approved by the Station Nuclear Safety Operating Committee.
2)
Station Nuclear Safety Operating Committee review and develop schedule to resolve significant preoperational test deficiencies and master deficiency list items acceptable to Region II.
3)
Complete development of operational procedures recommended by R.G. 1.33.
(Three procedures not approved.)
4)
Completion implementation of industrial security ' program in accordance with the North Anna Modified Amended Security Plan.
5)
Resolve Quality Assurance program outstanding items involving instrument calibratian control (79-11-13 and 79-11-24).
6)
Verify and all protective system instrument sensing lines are properly connected to their associated instruments (79-54-02).
b.
Item to'be Resolved Prior to Two Unit Operation in Modes 4 or Above Modify the electrical transfer bused tu eliminate overloading during two unit operation (LER 79-57/01T-0 for Unit 1; Unit 2 open item 79-28-10).
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
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vessel lower head insulation due to thermal growth during pre-critical testing (78-24-02).
2)
Visual verification of hydrculic snubber operability at operating temperature (79-26-02).
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Resolve lack of seismic supports and missile protection for the emergency diesel generator exhaust and mufflers (79-12-02).
4)
Demonstrate operability of safety related AC and DC solenoid valves which may have parts missing (79-21-02).
5)
Complete modifications to control room air conditioner chillers required as a result of the increase in projected service water temperature which resulted from reevaluation of the spray pond efficiency (79-21-01).
6)
Verify adequate seismic stress analysis of safety related piping per IE Bulletin 79-07 (79-28-07).
7)
Verify adequacy of service water and component cooling water pipe supports to resist thermal loadings over service water temperature range (78-36-01).
8)
Determine if atypical weld materialwas used in the reactor vessel seam welds (IE Bulletin 78-12A).
9)
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).
10)
Resolve error found in DNBR analysis for single dropped control rod (79-28-01).
11)
Resolution of potential degradation of electrical penetrations due to overcurrent conditions (79-46-01).
12)
Verify adequacy of emergency diesel generators fan cooling systems over all ambient temperature ranges (79-44-02).
13)
Replacement of cracked disc for containment isolation valve MOV-2380 (79-54-04).
14)
Upgrade safeguards building exhaust ventilation system to Seismic Class I (79-54-06).
15)
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).
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Verification that reactor trip breaker auxiliary contacts for
interlocks P-4 are adequately tested by solid state protection system surveillance procedures (79-56-03).
17)
Modification of circuitry to assure that engineered safety features equipment do not return to normal operating modes following reset of containment depressurization actuation (CDA) or safety injection (SI) (79-56-04).
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Items to be Resolved Prior to Exceeding 25% Power 1)
Develop surveillance program for fiberglass spray pond piping and supports (R. G.1.72).
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Complete environmental qualification testing and resolve any deficiencies identified for electrical components inside con-tainment (IE Circular 78-08/78-28-01 and IE Bulletin 79-01).
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