IR 05000338/1980038
| ML20003G596 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/05/1981 |
| From: | Kellogg P, Tattersall A, Webster E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20003G585 | List: |
| References | |
| 50-338-80-38, 50-339-80-36, NUDOCS 8104300226 | |
| Download: ML20003G596 (12) | |
Text
{{#Wiki_filter:8 % UNITED STATES
NUCLEAR REGULATORY COMMISSION o $ E REGION ll [[ 101 MARIETTA ST., N.W SUITE 3100 Q ATLANTA, GEORGIA 30303 ..... Report Nos. 50-338/80-38 and 50-339/80-36 Licensee: Virginia Electric and Power Company P. O. Box 26666 Richmond, VA 23261 Facility Name: North Anna Units 1 and 2 , . ' Docket Nos. 50-338 and 50-339 License Nos. NPF-4 & NPF-7 Inspection at North Anna Site near Mineral, VA j. Inspectors e e _ D /m Il E. H. Webger y{~ Da'te Signed 0% = =A K A}5}wI A. P. Taprsall 6[ Otte digned ~ Approved by: / ,b/ I ' P.
oco. Sodinn chief-Date Sighed / . . ~... - _ l SUMMARY - - Inspection on November 1-30, 1980 Unit 1 Areas Inspected - . , This routine inspection by.the resident inspectors involved 84 inspector hours on i ... site in the areas of task action plan items, licensee event reports, surveillance , i activities, previous inspection concerns and operational safety.
Unit 1 Findings Of the five areas inspected, no items of noncompliance oc deviations were identified in four areas. Two violations were identified in one area (Failure to complete a written safety review of a special test and Failure to test instrumentation in the frequency prescribed by Technical Specification.)
Unit 2 Areas Inspected This routine inspection by the resident inspectors involved 128 inspec' tor-hours on site in the areas of task action plan items, transient testing, plant trip, IE Bulletins, surveillance activities and operational safety.
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. . Unit 2 Findings Of the six areas inspected, no items of noncompliance or deviations were identi-fied in five areas.
Two violations were identified in one area (Failure to maintain required procedures current and Failure to complete required reviews of procedure changes).
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- . . . DETAILS 1.
Persons Contacted ~ Licensee Employees
- W. R. Cartwright, Station Manager
- E. W. Harrell, Assistant Station Manager
- D. L. Benson, Superintendent - Technical Services J. R. Harper, Superintendent - Maintenance i
S. L. Harvey, Superintendent - Operations ~ J. M. Mosticone, Operations Coordinator R. A. Bergquest, Instrumentation Supervisor J. P. Smith, Senior Engt.ieer .. P. T. Koutsen, Assistant Engineer ,. ' F. T. Termunella, Senior Engineer D. A. Heacock, Assistant Engineer A. Stall, Assistant Engineer Other licensee employees contacted included two technicians, four operators, two security force members, and seven office personnel.
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- Attended one or more exit interviews l
2.
Exit Interview .s . " "- The inspection scope and fin ~ dings were summarized on November 7, and' 21,- l with those persons indicated in paragraph 1 above. The violations described ! in paragraph 7 were discussed with plant management on November 21. The violations described in paragraph 8 were also discussed at that time. Plant management acknowledged each of the items, although they felt certain that a written safety analysis had been present during their review of the special test (ST)-32 procedure. 'No such analysis could be found, nor did any' i engineer on the staff. remember writing such an analysis.
i ... 3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items l Unresolved items were not identified during this inspection.
5.
TMI Action Plan (NUREG 0737) During this inspection period, NRC published NUREG 0737 " Clarification of TMI Action Plan Requirements", which clarified requirements published earlier in NUREG 0578, NUREG 0660, and several letters to all licensees, and changed requirement dates for completion of several tasks.
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Following an in-depth review of t;UREG 0737, the following action items for IE followup were identified for near term licensee action: a.
Item I.A.11 - verify the STA training program in place and all STA's on shift January 1, 1981 have completed the program.
(338/80-38-01 and 339/80-36-01) b.
Item I. A.2.3 and I. A.3.1 - verify the licensed operator training program in place and that training department instructors are involved (338/80-38-02 and 339/80-36-02).
c.
Item I.B.1.2 and I.C.5 complete the previously opened items in IE report 339/80-10: 80-10-01 through 80-10-07.
d.
Item I.C.6 - verify procedures are implemented to verify performance in ># accordance with ANSI N18.7-1972 Draft 3 ANS 3.2 January 1, 1981 (338/80-38-03 and 339/80-36-03).
e.
Item II.B.2 - verify plant shielding modifications installed as identi-fled in section 22.3 item II.B.2 of NUREG-0053 Supplement II for Unit 2 by January 1, 1981, (339/80-36-04) and in Unit 1, as identified in VEPC0 lettegs 306 dated April 1,1980, and 536 dated July 7,1980, by January 1, 1981 (338/80-38-04).
f.
Item II.E.1.1 - verify completion of upgrade of the emergency conden- - sate storage tank leve,l' indication circuitry to safty grade by _~ January 1, 1981. Design ~ changes DC 80-502 (Unit 1) and DC 80-542 (Unit ~ 2) have been developed'to complete this item and at the close of this inspection period the licensee reported completion of DC 80-542. These design changes will be reviewed when all documentation is completed (338/80-38-05 and 339/80-36-05).
- g.
Item II.E.4.1 - verify completion of remote operated isolation valves ' ' on the hydrogen analyzer and recombiner piping by January 1, 1981 ... . (338/80-38-C6 and 339/80-36-06).- 5.
Item II.F.1 - verify the wide range RCS pressure input to the sub-cooling meters is upgraded to safety grade by January 1, 1981 (338/80-38-07 and 339/80-36-07).
6.
Correctiva Action for IE Bulleting 79-27 Unit 2 License Condition 2.C.(12) required that prior to startup following the first outage of sufficient duration, or within six months from the date of issuance of this license, whichever comes first. VEPC0 shall implement the modifications related to IE Bulletin 79-27, "Los s of Non-Class IE Instrumentation and Control Power Systems Bus During Operation." VEPC0 . -- . . - . . -. --- . .. . * . .
letters 185 dated 3/7/80, 485 dated 3/30/80 and 602 dated 7/9/80, identified the following required changes: a.
Modify power supplies to the 3 loop wide range T hot, T cold loops from different vital buses: DC 80-552A.
b.
Provide manual transfer to alternate power supply for Main Control Board Vital Solenoid Operated Valve Panel A, Panel B, Vital Instrument Panel 2-I and 2-III: DC 80-554A.
c.
Provide manual transfer to alternate power supply.for the Main Control Board Annunciator System: DC 80-553A.
d.
Provide Main Control Board Indication for Semi-Vital Bus A,B: DC-80- , 556A.
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Provide DC powered indication for DC bus voltage on 125V DC 2-I, 2-II, 2-III, and 2-IV: DC 80-557.
During the November maintenance outage design changes (DC) 80-553A, 54A, and 56A, were completed on Unit 2.
These Design Changes were reviewed to determine completion status and implementation of commitments.
Upon completion of th,is review the inspector had no further questions on these changes. It should be noted that DC 80-552A and DC 80-57A are required to be completed by February 21, 1981, as stated in the referenced license , l condition. Completion of these_ Design Changes will be reviwed at a later , date and is designated as insp'e'ctor followup item 339/80-36-08.
- . . 7.
Valve Lineup Checkoffs During this inspection period, a review was made of the current Valve Lineup Checkoff s against the latest revision of the system flow diagram.
The following is a list of the Valve Lineup Checkoffs and flow diagrams: - - ~ 2-0P-7.1A (12/12/78) ~ Low Head Safety Infection System 2-OP-7.2A (5/16/79, Revision 1) High Head Safety Injection System 2-0P-7.3A (12/12/78) Safety Infection Accumulators 2-0P-7.4A (12/12/78) Quench Spray System 2-0P-7.5A (2/26/79, Revision 1) Outside Recirculation Spray System 2-0P-7.6A (12/12/78) Inside Recirculation Spray System 2-0P-7.7A (12/12/78) Refueling Water Storage Tank 2-0P-7.8A (12/12/78) NaOH Chemical Addition 2-0P-7.10A (3/5/80, Revision 1) Casing Cooling System i .. - --. _ ..-.. ,_ __.-__ _ - . __
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2-OP-21.9A (12/12/78) Control Room Bottled Air Pressurization System 2-0P-31.2A (7/23/80, Revision 2) Auxiliary Feedwater .
2-0P-46.4A (12/12/78) Diesel Air 2-0P-53.2A (12/12/78) Ems gency Generator Fuel Oil System , 12050-FM-96A-11, 96B-11, Safety Injection System ], 12050-FM-91A-11, 918-4, Containment Quench and Recirc Spray Sub Systems 11715-FB-34F-2, Compressed Dry Air Bottle System . .. ' 12050-FM-74A-12, Feedwater 11715-FB-35A-7, Yard-Fuel Oil Lines . 12050-FM-107A-2, Emergency Diesel Air Services The review consipled of comparing the valve checkoff and flow diagrams to see if all valves were shown on both the flow diagram and valve checkoff.
Where deficiencies were noted, the last completed valve lineup was reviewed to see if the discrepant condition was noted.
In addition, Procedure , Deviations to the above itsted Valve checkoffs were reviewed to see that the - ' required approvals were noted?' -- l . . l At the completion of this review, the discrepancies were combined under the following categories: j (1) Valves deleted from a system and removed from the flow diagram are still listed on the valve lineup and had been checked. (1 example), t ~ $ (2) Valves added to a system and to.the applicable flow diagram not added to the systems valve lineup checkoff (9 examples), (3) Valve lineups where Procedure Deviations had not been reviewed by two members df the plant staff, one holding an SR0 license or by the Station Nuclear Safety Operating Committee and approsad by the Station ' Manager (2 examples).
Technical Specification 6.8.1.a requires that written procedures be estab-lished, impl'emented, and maintianed covering activities described in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Paragraph 3 of Appendix A requires procedures for startup of systems as Emergency Core Cooling Systems and Auxiliary Feedwater Systems.
Items (1) and (2) demon-strate that these procedures have not been msintained up to date and are considered a Violation. (339/80-36-09).
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Technical Specification 6.8.3 requires that changes to procedures required by T.S.
6.8.1 may be made provided: (a) the intent of the original procedure is. not changed; (b) the change is approved by two (2) members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operators License on the Unit affected; (c) the change is documented, reviewed by the SNSOC and approved by the Station Manager within 14 days of implementation. Two deviations, as noted under Item (3), were not reviewed and approved as required in that the Procedure Deviation was made in April 1980, and not reviewed and approved until November 1980. This item is a Violation, (339/80-36-10).
~ - These items and the specific examples were discussed with Station Manage-ment. At this time, the following commitments were made: a.
Review the current valve lineups against the flow diagrams and -,. 7 completed Design Changes and correct noted discrepancies.
b.
Review completed valve lineups to determine if any further Procedure Deviations have not received required reviews. Any deficiencies noted - are to be corrected and action taken to insure the.;ystem lineup is correct.
i r c.
Increase emphasis on conduct of technical reviews upon completion of Design Changes so as to preclude similar' problems in the future.
, l. , ' maintenance shutdown and will be followed up in_ December (339/80-36-11) Item-Completion of Items a and-b. is_ required prior to startup from the current c above shall be followed up in a future inspection (339/80-36-12).
8.
RTD Cross Calibration Unit 1 Licensee Event Report 80-81/03L-0 reported that the calibration of the protection system resistance thermal detectors (RTD) - had not-been' r conducted at the specified eighteen (18)-month frequency - as required by .i . Technical Specification 4.3.1.1.1. Performance of the periodic test (PT) 1-PT-121 RTD Cross Calibration, which conducts this calibration i s not possible during power operations and therefore, an abbreviated form of this procedure was developed as a special test, 1-ST-32 RTD Cross Calibration at Power.
This test was conducted on September 25, 1980, and satisfactory results were achieved.
During further discussions concerning this report, the inspector determired that the operability of the RTD themselves which is the purpose of this cross calibration, appeared to have been adequately verified for these circumstances by the abbreviated procedure. Although there was comparison of three (3) RTD's instead of twenty-four (24) as called for in the-PT, there was sufficient assurance that the result was correct. Additionally, all electronic components of the associated protection system instrument , loops wera verified to have been calibrated in accordance with Techcnial Specification 4.3.1.1.1.
It was also determined during this review that the special test 1-ST-32 did not appear to have a specific Safety Analysis . .q.- -,, ,,w-p------ww--wyw---,,,-,.rg- ---9e-w-e.--e,,ur -'-w wa w4*e iV Y-4*W-P-*~-----+--v---r+c - + - - - - - e -*r--e --ww 7W1-'- n-e--w--v------*-Ne-a-7"-w-- --'*' M
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determining that this test does not constitute an unreviewed safety question.
This Safety Analysis is required by the Nuclear Power Station Quality Assurance Manual (NPSQAM) section 5.8.17 and 10 CFR 50.59b. Failure to have this safety analysis is a Violation, (339/80-38-08).
The inspector also determined during further review of' this event that the required calibration of the Post Accident Monitoring Wide Range Thot and Tcold RTD's had also not been conducted at the specified, eighteen (18) months, frequency as required by Technical Specification 4.2.2.6.
Failure to conduct this required surveillance is an violation (338/80-38-09).
Additionally, the need to revise LER 80-81 was discussed.
Those points needing further clarification in a supplemental report are: (1) the fact that a Special-Test was conducted to allow calibration at power, (2) that a review is currently underway to verify that all Periodic Tests are correctly .., / being scheduled. These comments were acknowledged by Station Management.
j 9.
Out of Phase Transfer of Emergency Buses Licensee Event Report (LER) 80-96/0IT-0, reported that the Architect / Engineer for the Sorth Anna Power Station had determined that under certain conditions a significant out of phase transfer could occur when the emer-gency buses werp transferred to the emergency diesel generators.
As interim corrective action, a two (2) second time delay was placed on closure of the Emergency Diesel Gencrator breaker to allow the bus residual voltage to collapse and prevent out of phase transfer. Final design and corrective . action will be completed when the engineering study is finalized.
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_, Installation of 'the additional two second time delay was conducted using Jumpers #415 for breaker 1H11, #416 for breaker IJ11, #206 for breaker 2H11 and #207 for breaker 2J11. The procedure for installation of the time delay and the required safety analysis were reviwed.
In addition, procedures 1-ST-34 and 2-ST-32 which were conducted to verify that the addition of the time delay functioned' as designed and that operation of the breaker'
manually, was not affected by this change, were reviewed.
There were no i , questions concerning the installation of the jumper and the required testing. This subject is designated as items 338/80-38-10 and 339/80-36-13 l and will be followed pending final resolution.
10.
Licensed Operator Disability . While observing the Startup Tests conducted on Unit 2, the inspector noted that the licensed reactor operator for Unit 1 had his right arm in a sling and the arm was bandaged.
Discussion with station management determined l that the operator had undergone minor surgery for an inflamation of the elbow joint. The inspector pointed out the requirement of 10CFR55.41 to ' report to the Director of Reactor Regulation within fifteen (15) days of its occurrence any disability, referred to in 10CFR55.11(a)(1), which occurs after the submission of his medical examination form. In that this surgery could affect the operators motor coordination, it appeared this should be reported.
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A short time later, the inspector reviewed a copy of a letter dated November 11, 1980, from the operator's physician, stating that the operator has suffered no motor impairment of his right arm. Station Management did, however, commit to submitting a report to the Director of Nuclear Reactor Regulation on this subject.
The licensee's report and promulgation of a reporting program shall be followed up in future inspections (338/80-38-11).
11.
Service Water Pump House Settlement Monitoring The inspector reviewed the surveillance results for the Service Water Dump House (SWPH) settlement monitoring for the year.
Periodic Test (PT) 113 requires a monthly survey of ten monitoring points in the SWPH and analysis of the overall settlement of the piping and differential settlement of the pump house to the piping.
,, /, Tha PT-113 data from January 1980 to October 1980, indicated the SWPH west side is still settling slightly faster than the east side, and that during the summer months this year, the overall settlement rate decreased signi-ficantly. As of October 1980, the SWPH had settled a total of 0.180 feet
from the baseline data ta(en in 1975, the service water piping, since its baseline in 1978, has settled 0.05 feet at about the same rate.
Technical Specif cation 3.7.12, table 3.7-5 identifies limits on overall piping settlement (0.22 feet), differential settlements between the piping upstream of the expansion joints and the SWPH (0.22 feet), the reservoir , pipe hangers and the SWPH-(0.17_ feet), and out of plane distortion of the
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~ SWPH (0.06 feet).
- In summary, this review verified that the licensee is closely monitoring SWPH settlement and that the data indicates the settlement rate has decreased, total and differential settlement are well within the prescribed limits, and the SWPH distortion is well within limits.
- . . In an attempt to further insure the accuracy of thsse surveys, the licensee i . has promulgated precedure PT-115, a surveillan e procedure for the contract surveyor. This procedure requies second order Class II surveying accuracy and has VEPC0 QC checks of instrumentation used in the survey.
The inspector had no further questions in this area.
12.
Plant Transient (Unit 2) On November 29,1980, Unit 2 returned to power operation following a twenty-six day maintenance outage.
At 10:30 p.m., several hours after reaching 100% power, the plant tripped on an electrical fault in the output from the main Senerator. At the close of this inspection period, the cause of the fault is not known, and licensee investigation into the scope of the damage is continuing. From log reviews, observations, and discussions with plant staff, shortly after the event, the inspector understood that a fault, somewhere between the main generator and the main transformers caused at least two phases of the generator output cabling to overheat and blow out - .
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potential transformers and conduit in the output bus work.
The main generator tripped on generator leads differential resulting in a turbine trip and reactor trip. It appears the reactor trip and subsequent shutdown operations were controlled and functioned as designed. The electrical bus transient did cause the two Unit 2 diesel generators to emergency start, but the transient was short enough that they did not load the emergency buses.
No emergency equipment was lost, no personnel infuries resulted, there were no fires following the event, although the turbine building filled with smoke and the onsite fire brigade responded.
Licensee investigation and repairs shall be followed up as the occur (339/80-36-14).
13.
Prompt Licensee Event Reports , ./ . ' The follwing event reports were forwarded by the licensee as prompt, or 14 day reports, as required by Technical Specification 6.9.1.8 or IE Bulletins.
The reports and pertinent licenseee actions were reviewed by the resident inspectors for completeness, completion of corrective actions specified, and technical adequacy of the corrective action.
Unit 1 LER 80.'84 dated October 16, 1980, Steamline Flow Setpoint Error.
a.
This item originated from a Westinghouse reanalysis of their " Pre-cautions, Limitations and Setpoints Document" (PLS), however, the licensee found that the Westinghouse recommendations forwarded in , , letter serial F5D/WGL-2651 dated October 2,1980, were also in error.
' ! Based on ifcensee engineering calculations mad; October 3, more - ' conservative steam flow summator gain and setpoint limits were inserted ! i on October 3-4, for both Units.
Renanlysis completed October 13, resulted in yet another correction to the summators, inserted on October 16, after comlete calculational checks were completed and ( confirmed by onsite Westinghouse staff and licensee engineering.
The inspector reviewed and checked the calculations used and verified the- - new settings, by review of completed. instrument procedures ICP-P-1-6.
" ^ P-446A, and 447, ICP-P-2-P-446. and 447.~ These procedures have been identified for revision to correct the summator alignment readings expected, and the interim procedure controls appear to be adequate.
l The generic questions of the Westinghouse error in the PLS document and any other plants affected has been referred to IE:RIV.
! b.
Unit 1 LER 80-91, dated November 12, 1980, Turbine Missile Damage Probability Increase Above 'FSAR Values.
This item resulted from i reanalyses by the archetect-engineer of turbine missile damage prob-l ability using revised turbine blade failure probabilities, developed by Westinghouse upon discovery of new blade cracking mechanisms. Westing-house analysis of the crack growth rate, the recent satisfactory Unit 2 turbine. inspection and the planned Unit 1 turbine inspectica in January 1981, assure turbine missile, damage probability to be less than . 2X10 -6.
The Unit 1 turbine inspection shall be followed up in future l inspections (38/80-38-12).
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. No violations or deviations were identified.in this area.
14.
Previous Inspector Concern - (Closed) Item (339/80-10-01) Operating Supervisor Operating License. At the time of the Unit 2 Management Inspection discussed in IE Report 339/80-10, the Superintendent-Operations did 'not hold a senior operators license on Unit 2.
Since that time, he has completed the NRC examination on Unit 2 and now has a senior operators license on both Units, Number SOP 3004-2, dated May 12, 1980.
~. 15. Plant Tours Tours of various plant areas were conducted during the inspection period in ,.2' conjunction with other inspection activites.
The following items, as available, were observed: a.
Fire Equipment Operability and evidence of periodic inspection of fire suppression equipment.
, , b.
Housekeeping Minimal accumulations of debris and maintenance of required cleanliness .
i levels in systems under at following testing. Observations regarding l certain areas were given~ to station management who acknowledged the--
inspector's comments. - c.
Equipment Preservation l Maintenance of special preservative measures for installed equipment as l I applicable.
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Component Tagging . Implementation and observance of equipment tagging for safety or l equipment protection.
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Communication
' Effectiveness of public system in all areas toured.
f.
Equipment Controls l Effectiveness of jurisdicational controls in precluding unauthorized work or systems turned over for initial operations or preoperational testing.
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Foreign Material Exclusion ' Maintenance of controls to assure systems which have been cleaned and flushed are not reopened to admit foreign material.
h.
Security Implementation of security provisions for both Units.
Within the above areas, no items of noncompliance or deviati6ns were observed when compared to the applicable station program and procedures.
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