IR 05000416/1993014

Insp Rept 50-416/93-14 on 930815-0918.Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint & Surveillance Observation,Preparations for Refueling,Employee Concerns & Reportable Occurrences

ML20059A409
Person / Time
Site:	Grand Gulf
Issue date:	10/05/1993
From:	Bernhard R, Cantrell F, Hughey C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20059A380	List: IR 05000416/1993014 ML20059A376 ML20059A387
References
50-416-93-14, NUDOCS 9310260324
Download: ML20059A409 (12)

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RECION ll 101 MARIETTA STREET, N.W., sUlTE 2900 ATLANTA, GEORGIA 30323 0199 e

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Report No.: 50-416/93-14 Licensee: Entergy Operations, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Nuclear Station Inspection Conducted: August 15, 1993, through September 18, 1993 Inspectors: .[kINIk/ _ff[5/F 3 R. ~ Bernhard, Ser>for Resident inspector Date Sfgned-fQ g' '] y } f gl}; 3 C. A. Hughey, Reside'nt' Inspector Date Signed Accompanying Per onnel- M. D. Sykes Approved by:

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IS. Cantrell, Chief / Date Signed Reactor Projects Section IB Division of Reactor Projects SUMMARY Scope:

The resident inspectors conducted routine inspections in the following areas:

operational safety verification, maintenance observation, surveillance observation, preparations for refueling, employee concerns programs, and reportable occurrences. Backthift inspections were conducted on August 17, 23, 24, and 25, and September 13, 15, and 16, 199 Results:

The licensee was aware of recent industry events involving Boraflex in spent fuel pool racks, and has periodically monitored Boraflex performance and integrity (Paragraph 3.b). Control room activities during a downpower, subsequent to a reactor scram, and a startup were observed. No discrepancies were indicated. Command and control and communications were good (Paragraph 3.c, g). ,

i The licensee's activities concerning the identification and replacement of an l eroded section of piping downstream of the reactor feed pump "A" minimum flow I valve were well planned and were completed satisfactorily (Paragraph 3.d).  !

9310260324 931008 PDR ADOCK 05000416 G PDR

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A violation for an inadequate procedure, which resulted in an inadvertent RCIC injection, was identified during backfilling the reactor level reference legs (Paragraph 3.h.(2)).

The inspector observed sludge being removed and transported from a settlement pond, which had been used for ten years, prior to performing an analysis for ,

radioactivity. Although subsequent analysis indicated no contamination, '

numerous industry events had resulted in the unauthorized release of radioactive materials (Paragraph 3.e). Delays were observed during maintenance activities due to the proper tools and equipment not being at job site when needed (Paragraph 4). There were three incidents involving failures on the post-accident sampling system within a five day period (paragraph 6a, b).

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REPORT DETAILS Persons Contacted Licensee Employees

• L. Daughtery, Superintendent, Plant Licensing W. Deck, Security Superintendent

• M. Dietrich, Manager, Training

• J. Dimmette, Manager, Performance and System Engineering

• C. Dugger, Manager, Plant Operations

• C. Ellsaesser, Technical Coordinator C. Hayes, Director, Quality Assurance

• C. Hicks, Operations Superintendent

C. Hutchinson, Vice President, Nuclear Operat* as

• Meisner, Director, Nuclear Safety and Regulatory Affairs

• D. Pace, General Manger, Plant Operations J. Roberts, Manager, Plant Maintenance

• R. Ruffin, Plant Licensing Specialist Other licensee employees contacted included superintendents, supervisors, technicians, operators, security force members, and office personne * Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragrap . Plant Status (71707)

A reactor scram from 100% power occurred on September 13, 1993, as a '

result of high reactor water level due to a spurious HPCS injectio This ended a continuous run of 403 consecutive days on lin Dan Pace was named General Manager, Plant Operations, on September 9, 1993. Mr. Pace was previously the Director of Nuclear Plant Engineerin ,

The licensee conducted a full scale exercise of their emergency preparedness program on August 25, 1993. The results of the exercise are discussed in NRC Inspection Report No. 50-416/93-1 . Operational Safety (71707 and 93702) Daily discussions were held with plant management and various members of the plant operating staff. The inspectors made frequent visits to the control room to review the status of equipment, alarms, effective LCOs, temporary alterations, instrument readings, and staffing. Discussions were held as appropriate to understand the significance of conditions observe !

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Plant tours were routinely conducted and included portions of the control building, turbine building, auxiliary building, radwaste building and outside areas. These observations included safety related tagout verifications, shift turnovers, sampling programs, housekeeping and general plant conditions. Additionally, the inspectors observed the status of fire protection equipment, the control of activities in progress, the problem identification systems, and the readiness of the onsite emergency response facilitie No deficiencies were identified.

b. The inspector reviewed the licensee's program for determining the ,

performance and integrity of Boraflex material in the spent fuel pool high density fuel storage racks. NRC Information Notice 93-70 discussed potential significant problems related to Boraflex degradation. The licensee was aware of recent industry events concerning Boraflex degradation. Two general methods were used to ,

determine Boraflex performance. During each refueling outage, discharged spent fuel was placed in preselected locations in the racks. After 10 to 14 months residence time, the fuel is removed .

to another location and a " Blackness Test" is performed. A fast neutron source / detector is dropped into the cells in a

" checkerboard" pattern to detect gaps in the Boraflex material as it shrinks and cracks from neutron exposure. This test had been performed 4 times with the most recent one being completed in May 1993. No significant cracking or gaps had been observed by the licensee. In addition, coupons installed with the racks are ,

removed and analyzed every five years for weight loss, boron content and material hardness. The last coupon removed in 1991 showed no significant signs of degradation. The inspectors concluded that the licensee was aware of recent industry events and the current program appeared to be adaquat ,

c. The inspector observed control room activities associated with a '

downpower conducted on August 23, 1993. Power was reduced from 100% to 70% power in accordance with Integrated Operated Instruction 03-1-01-2, Power Operations, Revision 33. No discrepancies were observed and the evolution was controlled very well. Control room command and control and communications were good. Power was reduced to lower general area dose rates in the condenser bay area prior to beginning Furmanite maintenance activities discussed below in paragraph ,

d. On August 21, 1993, during a routine control rod sequence exchange the licensee discovered severe erosion and cracking (greater than 50% circumferential) on the outlet reducer (8" to 12") of the non-safety related reactor feed pump "A" minimum flow valve (IN21-F503A). This valve discharges into the main condenser, a distance of about 5 feet. A significant amount of steam appeared to be leaking past the seat of IN21-F503A and impinging on the reduce Live steam could actually be observed through the crack moving through the pipe. After discovery, the "A" reactor feed pump minimum flow line was isolated by closing an upstream maintenance .

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block valve. This caused an increase in condenser inleakage and thus offgas flow. A temporary lead sheet was installed over the crack in an attempt to reduce inleakage with minimal results. On Agusut 23, 1993, a temporary support was installed just up stream of the minimum flow valve to support the weight of valve and to compensate for the reduced strength of the erroded joint. A temporary furmanite box (collar) was installed over the cracked portion of the piping. The inspectors viewed a videotape of the cracked piping. This piping was later replaced after an unplanned plant scram / forced shutdown occurred on September 13, 1993 (see paragraph 3.h). Repair activities for the minimum flow valve were scheduled for RF0 On August 31, 1993, the inspector observed sludge being removed from a drained settlement pond within the protected area of the pl ant. This pond had been used for approximately the last 10 years for the settlement and neutralization of makeup water demineralizer system regenerative wastes and rinse water prior to discharge. The sludge was being transported via dump truck to the unfinished Unit 2 cooling tower basin located outside the '

protected area but within the owner controlled are Investigations by the inspectors determined that a radiological analysis had not been performed on the sludge prior to remova A subsequent analysis performed by the licensee on September 3, 1993, indicated only naturally occurring radioisotopes. Although this pond was outside the radiological controlled area and non-contaminated, numerous industry events (NRC Information Notices 88-22, 92-11, and 91-40) have resulted in the unauthorized release of radioactive materials beyond restricted areas. The licensee agreed that increased awareness to these situations could significantly reduce the potential for these occurrences at Grand Gul The inspector reviewed the licensee's proposed hardware modifications to the Reactor Vessel Level indicating system in response to NRC Bulletin 93-03 and Generic Letter 92-04. The modifications were to be completed during RF06, which was to begin on October 8, 1993. This system was to provide a continuous backfill, using control rod drive pump discharge as a source of purge water, to the four reactor vessel level indicating system reference legs to prevent the buildup of noncondensible gase The installation and testing of this modification (Design Change Package 93-0011) will be followed as Inspector Followup Item 50-416/93-14-03. The planned hardware modifications appear to meet the intent of the bulleti The inspector observed control room activities associated with a power increase from approximately 1% to 4% on September 17, 1993, e and attended the Plant Safety Review Committee meeting where the on-shift scram analysis, open MNCRs, and applicable LCOs were review prior to start-up. No discrepancies were observed. Power was properly increased in accordance with Integrated Operating

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Instruction 03-1-01-1, Cold Shutdown To Generator Carrying Minimum Load, Revision 47. Mode 2 was entered at 7:29 a.m and criticality was reached at 11:40 a.m. with a period of 6 seconds. Recirculation loop A and 8 temperatures were 264 and 263 degrees Fahrenheit. The power increase was suspended at about 4%

due to an unknown source of condenser inleakage. This inleakage was estimated to be in excess of 200 cubic feet per minut Although aggressive attempts were underway to find the inleakage by employing helium leak detection methodology, the source of the inleakage had not been determined by the end of the inspection period and reactor power remained at 4%. The inspectors reviewed the activities associated with the events listed belo '

(1) On September 13, 1993, at 3:14 p.m., the high pressure core spray system (HPCS) initiated and injected into the reactor vessel due to several HPCS Low Water Level (Level 2) signal spikes below the setpoints on trip units B21NC021C and B21NC023G. HPCS injected for approximately 40 seconds with a maximum flow rate of approximately 4000 gallons per minute. The operators could not secure the injection prior to a reactor scram (from 100% power) occurring at 3:15 on high reactor water level (Level 8). No safety relief valves lifted and all safety systems functioned properly during the tri The inspector observed control room activities beginning about 10 minutes after the scram. By that time Integrated Operating Instruction 03-1-01-4, Scram Recovery, Rev. 27, had been entered. The inspectors observed a controlled scram recovery by the operations staff in accordance with procedural requirement The cause of the trip unit spikes that initiated the HPCS injection had not been determined by the licensee by the end of the inspection period. The completion of this evaluation will be followed as Inspector Followup Item 50-416/93-14-0 (2) On September 14, 1993, the day after the scram, maintenance personnel were in the process of manually backfilling the reference leg to reactor level condensing pot B210004B, which feeds shutdown cooling isolation instrumentatio This backfilling was performed as a result of the licensee's response to NRC Bulletin 93-03 prior to entering shutdown cooling until a permanent continuous backfill modification could be installed (see paragraph 3.f).

A level transmitter associated with this reference' leg (LT N0918) provided a signal to master trip unit B21N6918 (Reactor Vessel Level 1-ECCS Division II). Trip unit 821N6928 (Reactor Vessel Level 2-RCIC Initiation) was slaved ;

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to this unit. The master trip unit was placed in

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" CALIBRATION" prior to beginning work to prevent system initiations caused by pressure perturbations from the backfillin General Maintenance Instruction 07-S-13-60, Reactor Vessel Reference leg Purge, Revision 1, Attachment II, Step 7.7.9, specified 3 turns on the calibration unit to provide a signal above the setpoint of the trip units. This number was not suffirient to clear the setpoint of the i slaved unit (B21N6928) which made up half of the RCIC initiation logic. This slaved trip unit was located on a i rack separate from the master trip unit therefore the trip was not observed by the technicians performing this procedure. When backfilling operations commenced, perturbations in the reference leg caused trip unit B21N692F '

which was slaved to trip unit B21N691F, to tri This satisfied the other half of the RCIC initiation logic. RCIC injected for approximately 2 minutes at 800 gallons per minute before being terminated by the operators. Reactor level increased approximately 20 inches. This inadvertent RCIC initiation occurred due to an inadequate step in the

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work procedure which did not adequately take into account

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the setpoint of the slaved trip unit B21N6928. In addition, ;

master trip units B21N691F and slaved unit B21N692F were not specifically addressed in the procedure.

Technical Specification 6.8.1.c requires that written procedures be established, implemented, and maintained

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covering the applicable procedures recommended in Appendix ,

"A" of Regulatory Guide 1.33, Revision 2. Regulatory Guide 1.33 recommends that procedures for performing maintenance which can affect the performance of safety related equipment should be properly preplanned and performed in accordance with written procedures and documented instructions.

General Maintenance Instruction 07-S-13-60, Reactor Vessel Reference leg Purge, provided instructions for purging reference legs of noncondensible gases. Instruction 07-S-13-60 was inadequate in that it did not adequately address ,

the setpoint of the slaved trip units. This resulted in an :

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inadvertent RCIC injection on September 14, 1993. This incident was identified as Violation 50-416/93-14-0 One violation was identified for an inadequate procedure resulting in a '!

RCIC injectio . Maintenance Observation (62703) ,

During the report period, the inspectors observed portions of the T ,

maintenance activities listed below. The observations included a review of the MW0s and other related documents for adequacy; adherence to }

procedure, proper tagouts, technical specifications, quality controls, i and radiological controls; observation of work and/or retesting; and ,

specified retest requirement ;

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MWO DESCRIPTION 098227 Lube RHR "A" jockey pump bearing Clean and inspect RHR "A" room cooler fa Check coupling wear or alignment for RHR

"A" jockey pum Inspect slip rings and megger generator (HPCS diesel generator)

The inspectors observed several delays to the jobs listed above because the proper tools and equipment were not at the job sites when neede These delays resulted in unnecessary radiation exposure to maintenance personnel and additional outage tim Replace scram discharge volume level "C" transmitter (ICllN012C)

The proper personnel, tools and M&TE were available at the job site which resulted in a well planned and executed replacement in a minimum of time. The replacement transmitter was installed as Temporary Alteration 93/0013 because its capillary tube length was shorter that the original. The inspectors reviewed the alteration package and no discrepancies were observe Inspectors verified the modification authorized under Temporary Alteration 93/0012. This alteration provided for installation of a blank flange downstream of the Division 2 diesel generator jacket water standpipe drain valve IP75F0358 because of leakage past the seat of the normally closed IP75F0358 valve. While visually inspecting the modification, the inspectors noticed that the valve had a small amount of leakage past its packing. The inspectors discussed the situation with control room personnel to ensure that they were aware of the leakage. No operability concerns were observed by the inspector No violations or deviations were identified. The results of the observations in this area indicated that maintenance activities were effectiv . Surveillance Observation (61726)

The inspectors observed the performance of portions of the surveillances I listed below. The observations included a review of the procedures for j technical adequacy, conformance to Technical Specifications and LCOs; verification of test instrument calibration; observation of all or part of the actual surveillance; removal and return to service of the system or component; and review of the data for acceptability based upon the acceptance criteri l l

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06-0P-1Fil-V001, Rev. 24, Fuel Handling Platform Interlock Chec IC-lCll-R-2001, Rev. 22 Scram Discharge Volume High Water Level (RPS) Calibration (Channel C)

06-IC-1821-Q-1001, Rev. 20 Safety / Relief Valve High Pressure Trip / Low Low Set Relief /ECCS Vessel Injection Pressure Calibration No violations or deviations were identifie The knowledge level of the technicians performing the surveillances was very good. The observed surveillance tests were performed in a satisfactory manner and met the requirements of the Technical Specification . Reportable Occurrences (90712 and 92700)

The event reports listed below were reviewed to determine if the  ;

information provided met the NRC reporting requirements. The i determination included adequacy of event description, the corrective action taken or planned, the existence of potential generic problems and-the relative safety significance of each event. The inspectors used the NRC enforcement guidance to determine if the event met the criterion for licensee identified violation On August 20, 1993, at 12:13 p.m., the Post-Accident Sampling System (PASS) was declared inoperable to perform corrective maintenance on a leaking solenoid valve. This required isolation of instrument air thus rendering both liquid and gaseous sample paths inoperable. Repairs were completed and the system declared operable the same day at 5:13 p.m. The resident inspectors were notified and a one hour notification was made to the NRC Operations Center per 10 CFR 50.72(b)(1)(v).

On August 24,1993, at 9:45 a.m., the PASS was again declared inoperable to replace a valve that was leaking by its seat after water was observed by plant personnel . spraying from the dilution skid vent line to the turbine building ventilation syste Repairs were completed and the system declared operable on August 26, 1993, at 4:37 p.m. The resident inspectors were notified and a one hour notification was made to the NRC Operations Center per 10 CFR 50.72(b)(1)(v). The day before, repairs had been completed on two valves associated with the liquid dilution portion of the PASS. Although this failure did not require declaring both liquid and gaseous portions of the PASS inoperable, this was one of three incidents involving valve failures within a 5 day perio , On September 10, 1993, a buried telephone cable in the owner controlled area was severed accidentally by a trencher digging a ditch outside the protected area. The inspectors requested the control room check their Emergency Natification System (ENS) line

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8 after discovering that the resident inspector outside phone lines were inoperabl The control room subsequently determined the ENS to be out of service and declared it inoperable at 2:27 p.m. This cable provided phone service for the ENS, the Operational Hot Line to Claiborne County, the ability for outside calls into the plant, and the VIP 2000 offsite personnel notification syste Back u)

communications with the NRC Operations Center were established ay the control room via commercial lines through phone links to the Jackson corporate office. The operations staff periodically contacted the Operations Center until repairs were complete Service was restored to the ENS at 6:56 p.m. The resident inspectors were notified and a one hour notification was made to the NRC Operations Center per 10 CFR 50.72 (b)(1)(v). On September 13, 1993, at 3:15 p.m., a reactor scram occurred as a result of high reactor water level due to an inadvertent HPCS '

initiation (see paragraph 3.h.a). The resident inspectors were notified and a four hour notification was made to the NRC Operations Center per 10 CFR 50.72(b)(2)(ii).

No violations or deviations were identifie . Employee Concerns Programs (TI 2500/028)

Temporary Instruction (TI) 2500/028 was completed during the inspection period. The licensee had several avenues for employees to express concern The Ombudsman program was designed to specifically address nuclear safety and security concern In addition, this program operated outside normal chains of comman Lists of contacts were posted in the plant. The Quality Programs Department was also i structured to address safety concerns. In both programs, employees l expressing concerns could remain anonymou Response _to the Ombudsman l program had been minimal to none within the last three years. In I addition to the above programs, safety concerns expressed by exiting / terminating employees in interviews conducted by the personnel department were passed on to the Quality Programs Department for resolution. All three programs were procedurally controlle Employees were encouraged to first take safety concerns to immediate supervision for resolutio If the employee felt that the issue is not properly resolved, then they were encouraged to contact their Ombudsman ,

representative, Quality Programs, or the NRC Resident inspector / regional !

office. Employees interviewed by the inspectors were aware of these l programs. A discussion of these programs was also included in the l General Employee Training progra No violations or deviations were identifie .

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9 Preparation for Refueling (60705)

A total of 276 new fuel assemblies were received on site between August 25 and September 6, 199 During this period the inspectors periodically observed all steps of new fuel processing to 17sure that processing was completed in accordance with Performance and System Engineering Instruction (P&SE) 17-S-02-110, New Fuel Processing, Rev. These observations included removal of new assemblies from shipping crates, visual inspection of the assemblies for signs of damage and foreign debris, rod spacing verification, radiological surveys, and installation of new fuel channels. The inspector also observed the movement of new fuel assemblies between the shipping containers, inspection stand and spent fuel pool. Observed movements were tracked and documented per the applicable portions of P&SE Instruction 17-S-02-300, SNM Movement and Inventory Control, Rev. 8. Also, the applicable portions of P&SE Instruction 17-S-02-100, Criticality Rules, Rev. 2, were adhered to during processin The inspector noted that new fuel processing was very weli coordinate Several enhancements to the process were observed compared to the inspector's observations from the previous refueling outage. These enhancements included process improvements in the receipt of crated fuel and the transportation of these crates to the refueling floor, and the .

reduction in the distance and number of new fuel movements after being removed from the crate. These improvements helped the licensee complete new fuel processing approximately I week ahead of schedul Minor problems were observed with pens, clipboards and other loose items not being properly secured while over the spent fuel pool. These observations were relayed to cognizant licensee personnel for resolution. The inspector noted correction of these poor work practices during later observation No violations or deviations were identifie . Exit Interview The inspection scope and findings were summarized on September, 17, 1993, with those persons indicated in paragraph 1 above. Dissenting comments were not received from the licensee. 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:

l Item Number T_ype Description and Reference 50-416/93-14-01 VIO Inadequate procedure resulting in RCIC injection (Paragraph 3.h.b)

50-416/93-14-02 IFI Root cause of HPCS injection (Paragraph 3.h.a)

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50-416/93-I4-03 IFI Implementation of hardware modifications ~to Reactor Vessel Level Indicating system (Paragraph 3.f)

10. Acronyms and Initialisms DCP -

Design Change Package ECCS - Emergency Core Cooling System ENS -

Emergency Notification System .

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Engineering Safety Feature HPCS - High Pressure Core Spray 101 -

Integrated Operating Instruction LC0 -

Limiting Condition for Operation MNCR - Material Nonconformance Report M&TE - Measuring and Test Equipment MWO -

Maintenance Work Order '

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Nuclear Regulatory Commission P&SE - Performance and System Engineering RCIC - Reactor Core Isolation Cooling RF0 -

Refueling Outage SNM -

Special Nuclear Material i

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