IR 05000416/1987017

Insp Rept 50-416/87-17 on 870620-0717.No Violations or Deviations Noted.Major Areas Inspected:Licensee Action on Previous Enforcement Matters,Operational Safety Verification,Maint Observation & IE Bulletin Followup

ML20236H717
Person / Time
Site:	Grand Gulf
Issue date:	07/28/1987
From:	Butcher R, Dance H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20236H716	List: IR 05000416/1987017 ML20236H714
References
50-416-87-17, IEB-86-002, IEB-86-2, NUDOCS 8708050308
Download: ML20236H717 (9)

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UNITED STATES

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. o- NUCLEAR REGULATORY COMMISSION j f- 3g - -

7, o REGION 11 ' )

.g Ij 101 MARIETTA STREET, '* g . ATLANTA, GEORGI A 30323

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Report No.: 50-416/87-17 Licensee: ' System Energy Resources, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Nuclear Station

' Inspection-Conducted: June 20 thru July 17, 1987 Inspect r: ) W 7[M 7 .

tche'r, Senior Resident Inspector Date ign d Approved by: /t%M 72 H.C. Dance,T5ection Chief, Division Date Signed of Reactor Projects SUMMARY Scope: This routine inspection was conducted by the resident inspectors at the site in- the' areas of Licensee Action on Previous Enforcement Matters,

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Operational Safety Verification, Maintenance Observation, Surveillance 0bserva-

--tion,- Engineered Safety Features System Walkdown, Operating Reactor Events, Inspector Followup and Unresolved Items, IE Bulletin Followup, and Preparatio for Refuelin Results: No violations or deviations were identifie ,

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8700050308 870728 PDR ADOCK 05000416 ,

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REPORT DETAILS Licensee Employees Contacted J. E. Cross, GGNS Site Director C. R. Hutchinson, GGNS General Manager R..F. Rogers, Manager, Unit 1 Projects

• A. S. McCurdy, Manager,-Plant Operations J. D.' Bailey, Compliance Coordinator

• M. J. Wright, Manager, Plant Support

• L. F. Daughtery, Compliance Superintendent D. G. Cupstid, Start-up Supervisor R. H. McAnuity, Electrical Superintendent J. P. Dimmette, Manager, Plant Maintenance

• W. P. Harris,- Compliance Coordinator

.J. L. Robertson, Licensing Superintendent L. G. Temple, I & C Superintendent

'J. H. Mueller, Mechanical Superintendent L. B. Moulder, Operations Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent S. M. Feith, Director, QA

• F. W. Titus,~ Director, NPE

• C. C. Hayes, QA Supervisor Other licensee employees contacted included technicians, operators, security force members, and office personnel, Advanced Nuclear Fuels (ANF) Porsonnel t

C. Meyer, Quality Assuranc .

• Attended exit-interview Exit Interview (30703)

The inspection scope and findings were summarized on July 17, 1987, with those persons indicated in paragraph 1 above. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. The licensee had no comment on the following inspection findings:

416/87-17-01, Inspector Followup Ite "orrection of minor discrepancies noted during walkdown of the Combustible Gas Control System. (paracraph 7)

! .l Licensee Action on Previous Enforcement Matters (92702)

(Closed) Violation 416/84-51-0 The licensee issued Administrative Procedure 01-S-06-3, Control of Temporary Alterations, Revision 21 to clarify 'the audit requirement The inspectors checked the temporary alteration log, checked several active temporary alterations for accuracy ll

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and completeness, checked the most recent audit logs for accuracy and checked that the actions taken by the Operations Superintendent met the requirements of the subject procedure. No discrepancies were identifie (Closed) Violation 416/87-10-0 The licensee revised Administrative Procedure 01-S-06-3, Control of temporary alterations to clarify the requirements to revise an existing temporary alteration. The inspectors verified a detailed review of outstanding temporary alterations was accomplished and corrections made as necessar See Violation 416/84-51-01 closecut in this report for other inspection activities in this are , Operational Safety Verification (71707, 71709 and 71881)

The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant opera-tion Daily discussions were held with plant management and various members of the plant operating staff. The inspectors made frequent visits to the control room such that it was visited at least daily when an inspector was on sit Observations included instrument readings, setpoints and recordings, status of operating systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to l limiting conditions for operation, temporary alterations in effect, daily journals and data sheet entries, control room manning, and access controls. This inspection activity included numerous informal discussions with operators and their supervisor Weekly, when the inspectors were onsite, selected Engineered Safety Feature (ESF) systems were confirmed operabl The confirmation is made by verifying the following: Accessible valve flow path alignment, power supply breaker and fuse status, major component leakage, lubrication, cooling and general condition, and instrumentatio General plant tours were conducted on at least a biweekly basis. Portions of the control building, turbine building, auxiliary building and outside areas were visited. Observations included safety related tagout verifica-tions, shift turnover, sampling program, housekeeping at d general plant conditions, fire protection equipment, control of activitier in progress, problem identification systems, and containment isolation At least monthly, the licensee's onsite emergency response facilities + - toured to determine facility readines Monthly, the inspectors reviewed at least one Radiation Work Permit (RWP),

observed _ health physics management involvement and awareness of signifi-cant plant activities, and observed plant radiation control At least quarterly the inspectors reviewed the licensee's program to limit personnel radiation exposure As Low As Reasonably Achievable (ALARA).

Monthly, the inspectors verified licensee compliance with physical 1'

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security canning ~and. access control requirement At least quarterly i the inspectors verified the adequacy of physical security detection and as'sessment aid , No violations or deviations were ide'ntifie . -Maintenance Observation (62703)

During.the report period, the inspectors observed portions of the maintenance activities listed below. The observations included a i review of. the Maintenance Work Orders (MW0s) and other related documents ,

for adequacy, adherence to procedure, proper tagouts,. . adherence to

- technical . specifications, radiological controls, observation of all' or part of the actual work and/or retesting in progress, specified retest requirements, and adherence .to .the appropriate quality control Activities observed included:

MWO E72917, - Repla.;e Agastat control relay in the A Standby Gas Treatment syste MWO M73308, Rod 20-13 will not withdra MWO 173348, Division 2 level 1 trip unit B21-N091B failed channel chec q No violations or deviations were identifie '. Surveillance Observation-(61726) .

The' inspectors observed the performance .of portions of the surveillance'es-

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listed below. The observation included a review of the pro ~cedure for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual ,

-surveillance, removal from service and return to service of the system or j

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components affected, and review of the data for acceptability bared upon the acceptance criteria. Activities observed included:

06-ME-SP64-A-0017, Revision 2 Yard Fire Hose Hydrostatic Chec l 06-IC-1C11-M-0003, Revision 20, TCNs 1 and Scram Discharge Volume i High Water Level Float Switches (RPS) Calibration, j 06-0P-1011-V-0003, Revision 22, RWL Rod Block Functional Tes P-IP75-M-0002-02, Revision 32, TCNs 24 and 25. Stendby Diesel Generator (SDG) 12 Functional Test. The SDG functioned as designed ,

and would have performed its safety functio The inspector had I several couments < regarding the performance of the surveillanc These comments are as follows:

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In' general, the instruments on local panel 1H22-P401 are not labeled by number but they do have descriptive-title label Two Calibrated chart recorders were installed. One as required by j step ~ 4.6 of the ' procedure and another by a troubleshooting Mainte- !

nance Work Order (MWO). The recorders were installed prior to the j time the surveillance was started. Doors on control panels 1H22-P401 >

and 1H22-P115 were left unsecured to accommodate the recorder lead ;

.The inspector questioned if the control panels could be considered '

operable with the doors unsecure Operations is declaring the diesel generators inoperable during future surveillance until they get an engineering evaluation determining the effect of open doors on ,

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Step 5.2.5.c of the procedure directs system modifications by lifting i leads on one Termiaal Board (TB5-107) and jumpering sto another i Terminal Board (TB2-007) in control panel 1H22-P401. The Terminal' j Boards.were not permanently labeled and drawings were not used to ,

locate the terminal boards. Pencil marks on the panel structure was used as a reference to identify terminal boards. The licensee issued MWO's-to add labels to identify the terminal boards inside the noted ,

panel Step 5.2.26.a states the fuel oil transfer pump should automatically start when the. day tank level reaches 26 inches and should automat-ically stop when the day tank level reaches - 39 inche The fuel oil transfer pump started and stopped within one inch of the noted -

values. A tolerance of what is acceptable should be specified. ' the licensee will review this matte No violations or' deviations were identifie . Engineered Safety Features System Walkdown (71710)

AL complete walkdown was conducted on the accessible portions of the Combustible Gas Contrni System (CGCS). The walkdown consisted of an inspection and ' verification, where possible, of the required system valve alignment, including valve power available and valve locking ~where required, instrumentation valved in and functioning; electrical and instrumentation cabinets free from debris, loose materials, jumpers, evidence of rodents, and system free from other degrading conditions. The ,

system was found to be in a satisfactory condition and appeared ready to perform its safety function if called upon. The inspectors noted several minor discrepancies as discussed belo System Operating Instruction (501) 04-1-01-E61-1, Combustible Gas Control System, Revision 19, valve and switch lineup sheets specify valves E61-F009, F010, F012, F013, F056 and F057 should be open. These valves

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- should be normally closed valves as defined in the containment purge monitoring program described in SERI letter dated March 2,1987 (AECM-87/0039). The noted valves were actually close These valves may be !

opened as necessary for purging out should normally be close Valve position indication has been added in the control room for the j drywell purge compressor A/B discharge . check valves, E61-F002A and B; '

initial vacuum relief check valves E61-F001A and B and post LOCA vaccuum relief check valves E61-F004A and B. The 50I does not check these indicators to ensura operabilit Valve E61-F598D was mislabeled E61-F598 P&ID M-1091, Combustible Gas Centrol Systems, zone A3 references to containment cooling system drawing M-1108B (D-8). The correct reference should be M-1100 The inspector could not find identification number labels on the fallowing valves. E61-F025, F059, F027, FX026 and FX02 Control switch on panel 1H13-P842 for valve E61-F015 did not have the valve number on'the panel, only a descriptio Resolution of the above noted discrepancies will be tracked as Inspector Followup Item 416/87-17-0 No violations or deviations were identifie . Operating Reactor Events (93702)

The inspector reviewed activities associated with the below listed reactor events. The review included determination of cause, safety significance, performance of personnel and systems, and corrective actio The inspectors examined instrument recordings, computer printou6s, operations journal entries, scram reports and had discussions with operations, maintenance and engineering support personnel as appropriat ]

l On June 29, 1987, at 7:23 p.m. with the plant at 100% reactor thermal i power the operators received an offgas panel trouble alarm and the offgas j flow was indicating 80 SCFM and increasing rapidly. The main condenser vtcuum was decreasing so reactor power was reduced rapidly by reducing .

recirculation flow. Valve N62-F001B, the Stean Jet Air Ejector (SJAE)

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main steam supply valve, was observed to be closed and an operator was dispatched to take action to attempt to open the valve. Before the operator could take action the turbine tripped on low condenser vacuum (21 inches mercury) causing a reactor scra P'. ant conditions were stabilized. All systems performed as designed. The licensee's investiga-tion revealed that failure of a relay (N62-R33) in the control logic for l

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the ..SJAE ' main steam supply valve caused the valve . to close. A SJAE noncondensible gas removal valve N620F003B failed to go to the closed 1- position;and this left' the line from the' main condenser to the offgas system open and'the high flow recordeJ for the offgas system was from flow

'into the condenser.. The flow measurement instrument reads flow in either direction. A post trip analysis was conducted and concurrence for restart was'given on June 30, 1987. The failed relay was replaced under Mainte-nance Work 10rder 73240. The reactor was taken critical and the main generator was connected to the grid at 1:45 a.m. on July 2, 198 On July 6, 1987, at . 9:40 p.m. a Division 2 level 1 (-155 inches) master

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trip ' unit (821-N091B) was taken out of . service due to its deviation during channel checks. MWO 173348 was written to troubleshoot the transmitter and correct the- problem. The licensee initiated Limiting Conditions 'o r Operations (LCOs)' for affected systems which would give them .72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> L before entering a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ' hot shutdown action statement. Subsequently,

.at'9:00 a.m. on July 7,-1987, it was determined that the Load Shedding and Sequencing. (LSS) panel was also considered inoperable and TS 3.8. . states that .with one LSS : panel inoperable, the licensee has ' 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to-restore' the; inoperable panel to operable or be in at -least hot shutdown

. . - within" the' next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The licensee's 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> restoration time had

. ended i at 5:40 on July' 7, 198 The licensee prepared to start shutting down' and concurrently started restoring the Division 2 level 1 trip system. The restoration was' completed at 1:40 p.m. on July 7,198 .The licensee initiated Incident Report 87-7-5 to document the late determination that the LSS panel was inoperable. The NRC duty officer was called at 11:45 a.m. on July 7,1987, to notify the NRC of inadvertently entering 'a TS action statement. The Operations personnel on shift the evening of July 6, used Technical Support Instructions (TSI) 09-S-06-1, Technical Specification Instrumentation Loop Logic, to identify the affected systems on the loss.of trip unit B21-N091B. TSI 09-S-06-1 was deficient in that it did not identify the- LSS panel as being 'affecte i The licensee has issued a change to correct TSI 09-S-06- l

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On , July 7, 1987 at 0:30 a.m. an isolation signal for the Reactor Core

. Isolation Cooling (RCIC) system was received due to high steam flow signal {i from the'RCIC/ Residual Heat Removal (RHR) steam supply line. The licensee l had initiated MWO 173055 to reset the RCIC/RHR high steam flow isolation J trip setpoints to lower valves based on discovery of discrepancies in the I original GE calculations. TS Table 3.3.2-2 provides RCIC/RHR high steam l flow' isolation trip setpoints as $145 inches H2O (allowable s151 inches H20) but revised calculations indicate that these setpoints should. be a more- conservative value of 's37 inches H2O (allowable s43 inches H20).

The licensee initiated MWO 173055 to provide for the necessary rewor J

Following the rework and while restoring RCIC/RHR isolation valves, the ]

inboard isolation bypass valve E51-F076 was jogged open to permit warmup of the RCIC/RHR steam line. RCIC/RHR steam lime inboard isolation valve E51-F063 was shut as part of MWO I73055, but is a normally open ' valve.

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7 Valve E51-F076 was opened 'as it had been done -in previous evolutions but the new lower setpoints for high steam flow isolation resulted in an )

isolation signa The isolation was reset, the outboard steam line ]

isolation valve E51-F064 was reopened' and valve E51-F076 was slowly j reopened. Isolation valve E51-F063 was then opened. The RCIC system was -l then operated and functioned satisfactorily. The licensee is adding a

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caution note to slowly open bypass valve E51-F076 to prevent actuation of i

, the high steam flow isolation trip. Valves E51-F063 (inboard isolation l valve) and E51-F064 (outboard isolation valve) are normally open valves {

during plant operatio i No violations or deviations were identifie l Inspector Followup and Unresolved Items (92701)  !

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(Closed) Inspector Followup Item 416/84-51-0 The licensee completed

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l upgrading the legibility of all control room drawings at the end of the {

first refueling outage. Drawing accuracy and legibility are being tracked j by inspector followup. item 416/86-08-02 and the identification and control of root valves is being tracked by violation 416/86-20-0 This item is close (Closed) Inspector Followup Item 416/86-32-03. This event was addressed-in LER 86-032. The inspectors reviewed the licensee's corrective action Alarm Response Instructions 04-1-02-1H13-P864-1A-F4 and 04-1-02-1H13-P864-2A-F4 were revised to caution operations to ensure the internal'

input breaker is reset on inverters 1Y87,1Y88,1Y95 and 1Y96 after a loss of AC supply power to the inverter has occurred. Labels were added on the AC power supply breakers supplying the inverters to inform opera-tions to reset the inverter internal AC input breakers after power is re-established. The operator round sheets require the inverter breakers be checked each shif (Closed) Unresolved Item 416/86-32-01. The licensee has revised drawing M-1081A to identify manual valve C11-F431 as the isolation valve in the !

air line-to Scram Discharge Volume vent valve C11-F010A. System Operating ,

Instruction 04-1-01-C11-1 has been revised to show valve C11-F431 is a ,

normally locked open valve. The licensee was unaware of other similar condition No further action is require . IE Bulletin Followup (92703)

.(Closed) Bulletin 86-02, Static "0" Ring Differential Pressure Switche The licensee's response to Bulletin 86-02, AECM-86/0237 dated July 30, 1986, stated no -Static "0" Ring series 102 or 103 differential pressure switches were utilized as electrical equipment important to safety at GGNS. This item is closed.

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L 11 '. Preparation.for Refueling (60705)

In preparation for a planned refueling outage starting November 6,1987, new fuel started arriving on site on July 13, 1987. The licensee will be uceiving new fuel for the next few weeks ~for a total of 288 fuel bundle ' The inspector reviewed the licensee's procedures for radiation contamina-tion surveys for receipt of new fuel, inspection and storage of new fuel channels, receiving new fuel, inspection and channeling in the new fuel inspection stand and'the storage of new fuel. The licensee had specified clear lines of supervision, radiation control requirements, shift manning,.

training of key personnel, quality assurance --requirements and had a representative from the fuel vendo Advanced Nuclear Fuel (ANF), on site. The -inspector witnessed off-loading of- the crates from the truck, transfer of Reactor Assemblies (RA) containers from the turbine building railroad bay'to the 208 foot fuel handling floor, and transfer of fuel bundles'from the RA stand to the new fuel inspection stand. The new fuel was then visually inspected, channels installed and then transferred to the spent fuel pool for storag No violations or deviations were identifie j

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