IR 05000369/1986035

Insp Repts 50-369/86-35 & 50-370/86-35 on 861021-1120. Violation Noted:Operating Procedures OP/2/A/6100/01 & OP/2/A/6250/03A Not Properly Implemented During Unit 2 Startup

ML20215E931
Person / Time
Site:	McGuire, Mcguire
Issue date:	12/09/1986
From:	William Orders, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20215E911	List: IR 05000369/1986035 ML20215E909 ML20215E918
References
50-369-86-35, 50-370-86-35, NUDOCS 8612230176
Download: ML20215E931 (11)

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Sm negg . , UNITED STATES i b NUCLEAR REGULATORY COMMISSION

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REGION 11 I" g j  ? ! 101 MARIETTA STREET. * * I ATI.ANTA, GEORGI A 303:3 s,*..,*/ -

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Report Nos.: 50-369/86-35 and 50 3 0/86-35

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Licensee: Duke Power Company ,,

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422 South Church Street- -

Charlotte..NC 28242 Facility Name: McGuire Nuclear Station 1.and 2 Docket Nos.: 50-369 and 50-370 License Nos.: NPF-9 and NPF-17 Inspection Conducted: October 1 , November 20, 1986 Inspector: f//f dtsu hhx1)

W. Orders, Senidr Resident Inspector

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'Dpte'51gned Accompanying Personnel: S. Guenther Approved by: [

T. A. feebles,'Section Chief

/J./r/T4 D4te/ Signed Division of Reactor Projects SUMMARY Scope: This routine, unannounced inspection involved the areas of operations safety verification, surveillance testing, maintenance activities, and event follow-u Results: Of the areas inspected, one violation was identified in the area of facility operations.

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REPORT DETAILS

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Persons Contacted o'

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Licensee Employees

• T. McConnell, Plant Manager

• B. Travis, Superintendent of Operations

• D. Rains, Superintendent of Maintenance

• B. Hamilton, Superintendent of Technical Services '

• S. McInnis, Nuclear Production Engineer

• Sample, Superintendent of Integrated Scheduling E. McCraw, License and Compliance Engineer

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Other licensee employees ' contacted included construction craftsmen, technicians, operators, mechanics, security force members, and office personne * Attended exit intervie . Exit Interview The inspection scope and findings were summarized on November 24, 1986, with those persons indicated in paragraph 1 above. One Violation, one Unresolved Item, one Licensee Identified Violation, and one Inspector Follow-up Item were . discussed. The licensee acknowledged the findings and offered no dissenting comment Violation: 370/86-35-03, Failure to follow procedure resulting in contaminated spil Unresolved Item: 369,370/86-35-04, Possible inadequate safety review for KC temporary modificatio Licensee Identified Violation: 369,370/86-35-02, Inadequate document contro Inspector Follow-up Item: 370/86-35-01, Personnel error resulting in reactor tri The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio . Licensee Action on Previous Inspection Findings (Closed) Inspector Follow-Up Item (IFI) 370/86-30-03 - Verify valve 2-CF-23 stroke testing. The packing on Valve 2-CF-23 was adjusted in September while Unit 2 was operating at power, so it was not possible

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.to perform a full valvel stroke timing test as required by the -

licensee's valve testing program. Unit 2 shut down on _0ctober 28 and

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12-CF-23 successfully passed its stroke test on November 10.

I l(Closed)1IFI 370/86-28-05 - Verify the licensee's . inspection of the

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Unit 2 pressurizer code safety valve blowdown ring settings. The- ;

licensee inspected the Unit 2 code safety valves (2-NC-1, 2-NC-2, and a

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2-NC-3) during the Rotork Valve operability ' outage commenced on October 28 and ' verified that all three valves' blowdown rings were" ,

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,1 F .i; 4. - Unresolved Items .

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i An ' Unresolved Item is a matter about which more information is required to [

determine whether it is acceptable or may involve a violation or deviatio ..

l- One . unresolved item was identified during this report period and is

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discussed in paragraph , Plant Operations The inspection staff reviewed plant operations during the report period, to l verify conformance' with applicable regulatory requirements. Control room j -logs, shift supervisors' logs, shift turnover records and equipment removal and restoration records were routinely perused. Interviews'were conducted with plant operations, maintenance, chemistry, health physics, and performance personnel.

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Activities within the control room were monitored during shifts and at shift changes. Actions and/or activities observed were conducted as prescribed.in applicable station administrative directives. The complement of licensed personnel on each shift met or exceeded the minimum required by Technical Specifications.

l Plant tours taken during the reporting period included but were not limited '

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to the turbine buildings, auxiliary building, Units 1 and 2 electrical equipment rooms, Units ) and 2 cable spreading rooms, and the station yard l

r-Zone inside the protected crea,

! During the plant tours, ongoing activities, housekeeping, security, equipment status and radiation control practices were observe Unit 1 Operations The unit was on line at essentially 100 percent power for the period from October 21-29, 1986. A spurious power runback to about 50 percent occurred on October 21 during the performance of switchyard relay testin Power-circuit-breaker-8 (PCB-8) was tripped open as directed by the test '

procedure and should not, in and of itself, have caused a problem. However, l the digital electrohydraulic (DEH) control system already believed PCB-9,

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the redundant busline output breaker, to be open, and, therefore, initiated a turbine runback to reduce generator output to within the capability of a single operable busline. In actuality, PCB-9 relay contacts had been damaged resulting in a false open indication in the DEH logic. The unit responded normally during the runback and full power operation was promptly resume A controlled unit shutdown was commenced at about 9:00 p.m., on October 28, when a design evaluation of the Rotork valve actuators on 1-NV-244 and 1-NV-245 caused the licensee to be.lieve that tl.ere may be several valves incapable of performing their safety functions under accident condition Valves 1-NV-244 ar.d 1-NV-245 were immediately declared inoperable and placed in the closed (safety) position and the requirements of Technical Specifi-cation 3.0.3 were invoked. The unit was taken off line at 2:36 a.m., on October 29 and reached cold shutdown conditions (Mode 5) at 4:00 p.m.,

the same day. The unit remained in Mode 5 for the balance of the reporting period pending completion of the Rotork Valve actuator design evaluation and the necessary actuator adjustments. This issue is further discussed elsewhere in this report and in NRC Inspection Report Nos. 369,370/86-3 Unit 2 Operations The unit operated at full power from the beginning of the reporting period until the morning of October 28, 1986, when the licensee discovered that the 18-month Technical Specification (TS) surveillances on both trains of the containment annulus ventilation (VE) system had not been performed within the time period required by the TS. Both trains of VE were declared inoperable and a power reduction was commenced pursuant to TS 3.0.3. The load decrease was secured at about 1:30 p.m., when VE train "A" testing was complete Power was held at about 45 percent while evaluating the possibility of a further power reduction to enable replacement of a hydrogen igniter which had failed since the last unit startup. A load reduction to 10 percent was subsequently commenced at about 8:00 As discussed under Unit 1 Operations, at 9:00 p.m., on October 28, the licensee made the determination that several Rotork valve actuators could be incapable of cycling their associated valves under accident conditions. As on Unit 1, the charging header isolation valves (2-NV-244 and 2-NV-245) were declared inoperable and placed in their safety position (closed) and the requirements of TS 3.03 were invoke Unit 2 was taken off line at 10:46 p.m. and preparations were made for a plant cooldown; Mode 5 was reached on the afternoon of October 3 The Rotork Valve actuator operability issue is further discussed elsewhere in this report and in NRC Inspection Report Nos. 369,370/86-3 The unit was returned to service early on November 18 and had reached about 28 percent power when an electrical fault on the isolated phase bus (IPB)

caused the generator output breakers to open, which in turn caused a turbine runback. Reactor power was reduced by the runback and was eventually stabilized at about one percent af ter the turbine was manually tripped,

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l pending investigation and repair of the IPB fault. The licensee discovered that moisture had accumulated in the IPBs due to condensation and was able busline back in service on the evening of

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to dry and place the '.'B" November 18.' Power was restricted to 50 percent until the "A" busline was restored to service about midday on November 19. This transient is further discussed later in this report.

, At 10:14 a.m., on November 20, the unit tripped from full power when 6900

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volt bus 2TA, which supplies power to the 2A reactor' coolant pump, was inadvertently deenergized. All safety systems functioned normally during

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the trip and the unit was returned to service early on the morning of L November 21. The licensee's investigation of the loss of power to bus 2TA will' be reviewed and tracked as an Inspector Follow-up Item (IFI 370/86-35-01). .

6. Surveillance Testing l Selected surveillance tests were analyzed and/or witnessed by the inspector

! to ascertain procedural and performance adequacy and conformance with applicable Technical Specification Selected tests were witnessed to ascertain that current written approved procedures were available and in use, that test equipment in use was ;

calibrated, that test prerequisites were met, system restoration completed '

i and test results were adequate.

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Missed Technical Specification Surveillances l Three instances of missed Technical Specification required surveil-l lances were identified during this reporting period. They are briefly i

discussed belo '

l (1) .TS 4.0.5 specifies the surveillance requirements for inservice i inspection and testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 valves. The McGuire valve inservice '

l testing program ensures that TS requirements are satisfied by l

testing these valves as required by Section XI, Subsection IWV, of the ASME Boiler and Pressure Vessel Code,1980 Edition. In l accordance with TS 4.0.3, failure to perform a surveillance l requirement within the specified time interval constitutes a failure to meet the operability requirements for a limiting condition for operatio At about 2:00 p.m. , on October 14, 1986, Performance personnel l discovered that the nuclear service water (RN) valve stroke timing

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l l quarterly periodic test, PT/1/A/4403/02, had not been performed on t l valve IRN-16A, th suction isolation valve for RN pump 1A, within l the specified time interval. The allowed time interval had expired l at midnight on October 1 :

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g (2) TS 4.6.1.8 requires that 18-month carbon and HEPA filter performance tests be performed on the containment annulus ventilation system. On October 28, the licensee discovered that, due to an entry error, a computer program had incorrectly projected the latest date for the Unit 2 test as November 29, 1986. The correct latest date for the surveillance test should have been August 29, 1986. This incident is further discussed in the plant operations section of this repor (3) TS 3.8.1.1 requires that, with one offsite electrical transmission circuit inoperable, the remaining AC electrical sources be demonstrated operable within one hour and at least once per eight hours thereafter. When busline 2B was removed from service on November 18, 1986, as discussed in the plant operations section of this report, the required operability tests of the 2A and 2B emergency diesel generators were not performed within one hou These three examples of failure to perf-- rveillances as required constitute apparent violations of the accility's Technical Specifica-tions and would, under normal circumstances, be cited as suc Since NRC Inspection Report Nos. 369,370/86-28, issued on November 5, 1986, cited two examples of failure to perform TS-required surveillances, and the licensee has not yet responded to that violation, no Notice of violation will be proposed at this tim Diesel Generator Test Documentation On the afternoon of November 14, 1986, while reviewing the licensee's documentation of a recent test failure of the 2A emergency diesel generator (DG), the inspector noted that no Diesel Generator Test Failure sheet (Attachment 2 to Operations Management Procedure (OMP)

2-6, " Diesel Generator Logbook") had been filled out for start attempt number 489. A review of the parent OMP revealed that Revision 4, dated October 3,1986, had deleted the Olesel Generator Test Failure Sheet (old Attachment 2) and had revised the remaining attachments to the OMP, including the "D/G Test Summary" shee Further investigation on November 17, revealed that the revised OMP had been properly filed in the control room copy of the Operations Manual, but that the superseded "D/G Test Summary" sheets had not been removed from the Diesel Generator Logbooks (both diesels, both units) and replaced with the revised version. A review of the logbooks indicated that the IA, 18, 2A and 2B 0/Gs had been tested 4, 4, 17 and 9 times, respectively, since Revision 4 to OMP 2-6 was approved on October When the matter was brought to the licensee's attention on November 17, it was learned that a staff engineer had already discovered the same discrepancy and that corrective actions were underway to advise the Operations' shif t personnel of the revision to OMP 2-6 and to replace its superseded attachment Further discussions with licensee

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management during the exit interview, impressed upon them the NRC's concerns regarding a potential weakness in the licensee's administrative control of changes to other 0MPs, Station Directives, etc. Licensee management present at the exit meeting agreed to evaluate the method of processing changes to those documents, and to establish administrative controls as necessar Criterion VI (Document Control) of 10 CFR 50, Appendix B, states that measures should be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. Although the administrative deficiency described above constitutes an apparent violation of this Criterion, a Notice of Violation will not be issued as permitted by 10 CFR 2, Appendix C. This issue and the licensee's corrective' action to prevent recurrence will be tracked as a Licensee Identified Violation (LIV 369,370/86-35-02). Event Follow-Up Unit 2 Turbine Runback on November 18, 1986 At the time of the transient, the unit was operating at about 28 percent power with the 2B main feedwater (CF) pump in service and the

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2A pump in standby (at reduced speed and not pumping). Shortly after the unit runback at 4:38 a.m., the 2B CF pump tripped on high discharge pressure and steam generator water levels began to decrease because of

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the loss of feedwater input. To aid in steam generator water level recovery and prevent the reactor from tripping on a low water level signal, the operators manually started both motor-driven auxiliary feedwater (CA) pumps and proceeded to place the 2A CF pump in servic While reviewing the transient, the inspector questioned whether the NRC Operations Center had been notified of the manual engineered safety feature (i.e., CA) actuation pursuant to 10 CFR 50.72. The senior reactor operator on duty stated that the notification criteria had been evaluated and it was decided that no notification was necessar The inspectors later consulted with the Region II staff and confirmed that the licensee had erred in not making the required notification Further discussions with the licensee's staff resulted in their making the required four-hour report (10 CFR 50.72(b)(2)(ii)) on the afternoon of November 1 NRC Inspection Report Nos. 369,370/86-28 issued on November 5,1986, cited two examples of failure to make required NRC notifications pursuant to 10 CFR 50.72. Since the licensee has not yet responded to that violation, nor had time to complete the appropriate corrective

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, Mechanical Penetration Room Spill Unit 2 was in hot standby (Mode 3) on the morning of November 17, 1986, l with a unit heatup in progress in accordance with OP/2/A/6100/01, the

" Controlling Pror.edure for Unit Startup". At approximately 5:00 a.m.,

the Operations staff discovered that the steam generator wet layup recirculating pump (BW) discharge relief valves were lifting and had  !

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spilled an estimated 3000 gallons of steam generator water to the 716 '

l ft. elevation mechanical penetration room. The operators isolated the spill and secured the BW system in accordance with Enclosure 4.2 of OP/2/A/6250/03A, " Steam Generator Cold Wet Layup Recirculation."

A review of the incident revealed that an operator had initialed Step 12 of the " Pre-heatup Checklist" (Enclosure 4.2) of OP/2/A/6100/01 as complete, indicating that no-load programmed Steam Generator (S/G)

levels had been previously established per OP/2/A/6250/03A. In actuality, the no-load S/G 1evels were established earlier in the

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week under a different procedure; Enclosure 4.1, " Draining S/G for l Inspection or Maintenance", of OP/2/A/6250/038, " Steam Generator", had  !

been performed on November 15, 1986, as a chemistry control measur The BW System had never been properly secured in accordance with l OP/2/A/6250/03A prior to commencing the plant heatup and pressurization i on November 17. The operator had erred in initialing Step 12 of the

" Pre-Heatup Checklist" as complete without actually verifying that the procedure referenced in that step had been performe Consequently, l the BW System was pressurized to its relief valve setpoint and a significant amount of secondary coolant was spilled to the auxiliary butiding floor exacerbating existing contamination problem This operator error constitutes an apparent violation of Technical Specification 6.8.1, in that OP/2/A/6100/01 and OP/2/A/6250/03A were not properly implemented during the Unit 2 startup performed on, or about, November 17, 1986 (370/86-35-03).

8. Rotork Valvt Actuators During a follow-up engineering / design evaluation performed pursuant to Inspection and Enforcement Bulletin (IEB) 85-03, Duke Power Company (DPC)

discovered problems with valves operated by Rotork valve actuators at the ,

MCGuire, Catawba and Oconee Nuclear Stations. Specifically, the problem involved valves for which the factory-set torque switch settings had been changed using a vendor supplied generic correlation between actuator torque ,

output and torque switch setting. This could have caused valve actuator '

motors to de-energize before the valves completed their trave l Based on this information and field analysis of installed equipment, the

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licensee determined that certain safety related valves may not be able to perform their safety related function under all condition Consequently, i at 9:00 p.m. on October 28, both McGuire units were shut dow '

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The licensee entered an extensive program to evaluate and correct, if necessary, the valves in question. All Rotork actuated valves required for safe operation, numbering approximately 200 on McGuire Unit 2, were evaluated and placed in the following categories: Valves Acceptable "As Is": Valves Set on Max Torque - No change require Operability Statement - Potential for valve to not perform does not affect plant response,

, Valves with Torque Less than Max - Actual required valve torque compared to design valve torque and evaluated against Rotork worst case performance curv No change require Non-Safety -

Valves deleted from evaluation because it was determined they did not receive 1E power and had no safety function, Other - Valves placed in safety position and power remove I Valves with Reset Torque Switches:

Actual required torque based on actual differential pressure was compared to design torque and evaluated against worst case performance curve. Torque switch adjustment require III. Valves Requiring Test:

These valves were judged to have limitations on maximum applied torque and required testing to a specific value by the following methods, Bench Test - Operator set to a specific value on Rotork test benc Differential Pressure Test - Specific Valve Operator combination tested against actual required differential pressur Strain Gage Test - Operator performance set in place on valve by use of strain gage The resident inspection staff has reviewed and continues to monitor the inspection and testing of the valve For specific dotatis related to the above, refer to NRC Inspection Report Nos. 369,370/86-3 ___

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9 Maintenance Observations Rautine maintenance activities were reviewed or witnessed bj the resident inspectors to ascertain procedural and performance adequacy and conformance with applicable Technical Specifications. The selected activities witnessed were examined to ascertain that, where applicable, current written approved procedures were available and in use, that prerequisites were met, equipment restoration completed and maintenance results were adequate.

1 Component Cooling Water Surge Tank Temporary Modification On October 23, 1986, the inspectors noted that the manway covers on the Units 1 and 2 component cooling water (KC) surge tanks had been loosened and slid over, leaving the tanks continuously vented to the auxiliary building atmosphere. Polyethylene sheeting had been draped over the manway openings to prevent foreign material from entering the tanks and temporary modification tags (numbered 3954 and 3955 for Units 1 and 2, respectively)

were attached. An investigation revealed that the manway covers had been set ajar on October 25, 1984, in response to a 10 CFR 21 notification made by Westinghouse Electric Corporation on July 12, 1984. Westinghouse had identified a substantial safety hazard involving the possible overpressuri-zation and rupture of the KC surge tank Duke Power Company (DPC)

evaluated the 10 CFR 21 report, found it applicable to the McGuire facility, and took temporary compensatory action by venting the KC surge tanks via their manway cover A description of the postulated overpressurization scenario and the associated OPC safety analysis was presented in Licensee Event Report Number 369/84-2 The temporary modifications of the Units 1 and 2 KC surge tanks were installed under Work Request (WR) Nos. 85768 and 85769, respectively. The licensee evaluated these WRs for safety significance in accordance with Enclosure 13.1 of Maintenance Procedure MP/0/A/7650/70, " Control of Temporary Modifications", and decided that they did not increase the consequences of a malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report (FSAR),

Section 11.4.2.1.5 of the MCGuire FSAR describes the KC system as a closed fluid system which provides a positive means for preventing radioactive corrosion or fission products from being released from the station in the event of malfunction or failure of the various heat exchangers containing primary reactor coolan To preclude the release of volatile fission products in the event of cooler leakage, an interlock from a radiation monitor closes a valve that normally vents the KC surge tank to the auxiliary building atmospher The temporary modification of the KC Surge tanks on October 25, 1984, effectively defeated the purpose of the surge tank vent valve isolation on high radiation levels in the KC wate If a reactor coolant heat exchanger failure were to occur, any dissolved radioactive fission or corrosion

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h products' wouldi escape to the auxiliary building - atmosphere, thereby apparently increasing the consequences - of a malfunction of equipment

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). important to safety previously evaluated in the FSA . The licensee is currently evaluating the level of overpressure protection necessary for the KC ~ system and various permanent design modification options available- to : incorporate that protectio NRC Region II is evaluating the adequacy of the safety evaluation performed by DPC prior to '

implementing the temporary modifications and will track this issue as an Unresolved Item (369,370/86-35-04).

- 11. Follow-up on Service Information Letter (SIL)

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As requested by Region II management, the resident' inspector staff reviewed l General Electric SIL 445 entitled " Intermediate Range Monitor (IRM) Fuse Failure" dated July 26, 1986, and discussed the matter with the license The licensee is evaluating the SIL's applicability to the McGuire IRM

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