IR 05000416/1988019

Insp Rept 50-416/88-19 on 880820-0923.Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint & Surveillance Observation,Ros,Operating Reactor Events & Followup & Unresolved Items

ML20204E995
Person / Time
Site:	Grand Gulf
Issue date:	10/05/1988
From:	Dance H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20204E979	List: IR 05000416/1988019 ML20204E973 ML20204E981
References
50-416-88-19, NUDOCS 8810210486
Download: ML20204E995 (10)

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'4 UNITED STATES

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E NUCLEAR REGULATORY COMMISSION o REGION 11

101 MARIETTA ST., .. .. ATLANTA. OEORGIA 30323 Report No.: 50-416/88-19 Licensee: System Energy Resources, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 :

Facility Name: Grand Gulf Nuclear Station InspectionConupted: August 20 - S ptember 23, 1988 Inspector '

% N Y J. L . . this, (esident Inspector Dete /51gned Approved by: C Mh fi. C. Dance, Section Chief O/Iff[

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Division of Reactor Projects

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SU4tARY Scope: This routine inspection was conducted by the resident inspectors at '

the site in the areas of Operational Safety Verification, Maintenance Observa-tion, Surveillance Observation, Reportable Occurrences, Operating Reactor Events, and Inspector Followup and Unresolved Item .

Results: The inspector identified the following strength associated with the SERI organization in resolving problem Management involvement from first line superv!sor to site director exemplified sensitivity of management to resolva the problem with the i cracked diffuser plate on Divisions 1 and 2 Diesel Generator. Nuclear l Plant Engineering thoroughness and completeness of engineering

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evaluation was timel The following weakness was identified:

h SERI has implemented several initiatives in the areas of plant AlthoukonaimedatachievingnoscramsorESFactuationstherestillexist operat problems in procedure adherenc Within the areas inspected, the following violation was identified: Failure t to follow procedure on protective tagging which ultimately lead to a reacter scram, (paragraph 7).

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8810210486 861006

{DR ADOCK 03000416 PDC I

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REPORT DETAILS 1 Persons Contacted Licensee Employees

,1. G. Cesare, Director, Nuclear Licensing W. T. Cottle, GGNS Site Director D. G. Cupstid, Superintendent, Technical Support L. F. Daughtery, Compliance Supervisor J. P. Dimmette Manager, Plant Maintenance S.M.Feith,Olrector,QualityPrograms

• R. From, Manager, Emergency Preparedness

• C. Hayes, Quality Program Supervisor C. R. Hutchinson, GGNS General Manager Electrical Superintendent R. H. McAnulty, Technical Asst., Plant Operations Manager S.Moulder, L. McCurdy, Operations Superintendent J. H. Mueller, Mechanical Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent

• J.L. Robertson, Superintendent, Plant Licensing R. F. Rogers, Manager, Special Projects

• J. Summers, Compliance Coordinator S. F. Tanner, Manager, Quality Services L. G. Temple, I & C Superintendent F. W. Titus, Director, Nuclear Plant Engineering

• H. J. Wright, Manager, Plant Support J.W. Yelverton, Manager, Plant Operations Other licensee employees contacted included technicians, operators, security force members, and office personne * Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragrap . Licensee Action on Previous Enforcement Matters (92702)

Not inspected this report perio . Operational Safety, Radiological Protection and Physical Security Verifi-cation (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-tions. Daily discussions were held with plant management and various eembers of the plant aperating staff, t

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2 i The inspectors made frequent visits to the control room such that it was !

visited at least daily when an inspector was on site. Observations >

included instrument readings, setpoints and recordings status of operat-ing systems, tags and clearances on equipmer.t controls and switches, !

annunciator alarms, adherence to limiting conditions for operation, L temporary alterations in effect, daily journals and data sheet entries, control room manning, and access coritrols. This inspection activity included numerous informal discussions with operators and their super-visor l

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Weekly, when the inspectors were onsite, selected Engineered Safety l Feature (ESF) systems were confirmed operable. The confirmation is made i by verifying the following: Accessible valve flow path alignment, sower :

supply breaker and fuse status, major component leakage, lubricat'on, cooling and general condition, and instrumentatio l General plant tours were conducted on at least a biweekly basi Portions I of the control building, turbine building, auxiliary building and outside l areas were visited. Observations included safety related tagout veriff- l cations, shif t turnover, sampling program, housekeeping and general plant l conditions, fire protection equipment, control of activities in progress, '

and containment isolation. The licensee's i problem identification onsite emergency responsesystems}lities fac were toured to determine facility l readines The inspectors reviewed at least one Radiation Work Permit (RWP), observed l health physics management involvement and awareness of significant plant I activities, and observed plant radiation controls. The inspectors !

verified licensee compliance with physical security manning and access l control requirements. Periodically the inspectors verified the adequacy -

of physical security detection and assessment aid No violations or deviations were identified, i

4. Hafntenance Observation (62703) -

During the report period, the inspectors observed portions of the maintenance activities listed below. The observations included a review of the Hafntenance 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, t.nd adherence to the appropriate quality control WO 181636, Replace Temperature Thermocouple 204A on Division 1 Diesel Generato WV 183822, Rework / Troubleshoot Fisher 546 Electro Pneumatic Senso MWO M33295, Replace Guide Bolts per WI&IR on Division 1 Diesel Generato , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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MWOM82486,AdjustPackingonCR0IsolationValv HWO 182979, Replace N0298 Temperature Element for Lube Oil Outle MWO M84200, Investigate Jacket Water Leakage Into Engine Intercooler for Turbo Charge No violations or deviations were identifie . SurveillanceObservation(61726)

The inspectors observed the performance of portions of the surveillances listed below. The observation included a review of the procedure for technical adequacy, conformance to Technical Specifications, verification of test instrument calibration, observation of all or part of the actual surveillances, rewoval from service and return to service of the system or components affected, and review of the data for acceptability based upon the acceptance criteri IC-1E30-M-0001, Revision 21, Suppression Pool Makeup Time Delay Relay Calibration and Functional Tes IC-1821-M-1001, Revision 25, Safety Relief Valve High Pressure Trip / Low Low Set Relief /ECCS Vessel Pressure Injection Permissive Functional Test Channel P-1C71-H-0001, Revision 23, MSIV Closure RPS Functional Tes IC-1821-N-1012, Revision 24. ATWS Reactor Vessel Level / Reactor Pressure Functional Tes IC-1011-H-0003 Revision 21 Scram Discharge Volume High Water Level FloatSwitchesCalibratio No violations or deviations were identifie . Reportable Occurrences (90712 & 92700)

The event reports listed below were reviewed to determine if the informa-tion provided met the NRC reporting requirements. The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event. Additional inplant reviews and discussions with plant personnel as appropriate were conducted for the reports indicated by an asterisk. The event reports were reviewed using the guidance of the general policy and procecure for NRC enforcement actions, regarding Ifcensee identified violation . _ _ _ _ _ _ _ _ _ _ .

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LER N Event Date Event '

88-012-00 August 15, 1988 Reactor Scram Induced by Lightening Strikes Affecting Neutron Monitoring Syste The event of LER 88-12-00 are still being evaluated and corrective actions will be provided in a followup LE , i No violations or deviations were identifie . OperatingReactorEvents(93702) ,

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The inspectors reviewed activities associated with the reactor events l listed below. The review included determination of cause safety signi- ,

ficance, performance of personnel and systems, and corrective action. The inspectors examined instrument recordin computer printouts, operations journal entries, scram reports and hahsdiscussions with operations, maintenance and engineering support personnel as apprcpriat On August 30, 1988, at approximately 2:05 a.m., the licensee received an ,

unexpected 1/2 scram on Channel C during performance of Surveillance 1 Procedure 06-0P-1C71-M-0001, MSIV Closure RPS Functional Test, when the i handswitch for MSIV B21-f0228 was returned to the auto position after

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being in the test position. The 1/2 scram was cleared in accordance with ,

the procedure. An attempt by the licensee to repeat the step twice, to I evaluate the unexpected 1/2 scram, was unsuccessfu MWO 183986 was  ;

written to investigate the unexpected 1/2 scram.

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On September 5,1988, during performance of Surveillance Procedure,  :

06-0P-SP64-(-0009, Fire Protection System Quarterly Valve Testing, Valve  !

IP64F2828 d d not meet the required stroke time. LCO 88-752 was written in response to the val * not meeting Technical Specification stroke time i

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of less than 4 secone' . The actual time was 5.6 seconds. MWO M84063 was written to clean the exhaust valve on the actuator. A tagout was written  !

to deactivate valve IP64F282A which is the companion valve 11 the pene- t tration. The tagout had been written by an operator and approved by his  !

supervisor. Valve IP64F282A was closed and deactivated by removing its control power fuse. The tagout also required the operator to open breaker  !

52-1P53113 to valve IP64-F282A control solenoid and breaker 52-115134 to valve P64-FA10A control solenoid. 501 04-5-01-P64-1, Revision 23, was  ;

used by the control room operator as a reference document to determine the l proper breaker for the valve tagouts without investigating all loads .

affected by the power panel breaker. When the operator opened breaker 1 52-1P53113, power was lost to valves IP53-F026A (Division 1 PSW) and [

IP53-F001 (Instrument Air) in addition to P64-F282A. The closure of l IP53-F001 isolated instrument air to the auxiliary and containment i building. Several alarms were received on panel P870 whea the breaker was  !

opene Several unseccessful attempts were made by the operating shift by i

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radio and PA system to reach the operator so that he could reclose the

breake The operating shif t noticed multiple control rod drifts were

' alarming due to lost of air to the scram valve pilot air heade The shif t superintendent gave directions to scram the reactor because of the multiple rod drifts at approximately 6:30 p.m. per the Off Nonnal Event Procedure (ONEP). Review by the licensee of GETARs data NSSS and BOP SOE i

logsshowedthattheplantreceivedanautomaticscramsIgnalandscrammed on scram discharge volume high level approximately 2 or 3 seconds before

, the manual scram signal occurred, The plant responded as expected with no ECCS being initiated, the recirculation pump did down shift to slow speed on low water level, reactor vessel water level shrunk to 10 inches and swelled to 48 inches before stabilizing out at approximately 36 inche Incident Report 88-9-1 was initiated to document this event by the license On June 30, 1987 a similar incident occurred causing closure of seventeen air-operated isolation valves. This event was reported by the licensee pursuant to 10 CFR 50.73 (a)(2)(iv) in LER 87-10. Technical Specification 6.8.1.a requires that procedures be established, implemented and main-tained covering the applicable procedures recemmended in Appendix "A" of Re ulatory Guide 1.33, Revision 2, dated February 1978. Appendix "A" of Re ulatory Guide 1.33, Section 1.c, states that administrative procedures sh uld cover equipment control (e.g., locking, and tagging). The licensee Administrative Procedure 01-5-06-L, Revision 21, Protective Tagging System, provided precautionary measures for 120 Vac power panel breakers in that because these power panel breakers normally provide control power to multiple components, when tagging these breakers, the shift supervisor /

superintendent should evaluate the effects on plant operation. Operating aids to assist in this evaluation have not been provided by the license Contrary to the above, the licensee failed to follow procedure on protec-tive tagging for evaluating whether opening breaker 52-IP3113 had any affect on other equipment and plani, operation. This will be identified as Violation 416/88-19-0 At approximately 4:00 a.m., on September 8,1988, drywell leakage (Floor Drain) increased to 4.25 gpm. The licensee commenced reactor shut down at 3:45 a.m., to less than St. power to investigate the incresse in unidenti-fled leakage. LCO 88-769 was entered approximately 12:00 (noon) when leakage had exceeded 5 gpm per Technical Specification requirement 3.4.3.2.b. The leakage had decreased within the Technical Specification limits at approximately 4:00 p.m. and the LCO was cleared. When reactor power was reduced to approximately 3.5% at 4:00 p.m., the first team entered the drywell to investigate the leakage problem. The team observed an approximate 8 f t. to 10 f t plume from a drywell equipment sump stub tube. Several attempts were made to isolate the leakage by torquing down on several manual drain / vent valves (F025A, F026A,..etc) with no succes The stub tube stands approximately 1.5 feet above the drywell floor and is routed through the 0/W floor to the equipment sump. The top portion above

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the floor is normally capped off, however, the team found the threaded cap of The side of the stub tube above the drywell floor is penetrated by vent / drain lines from other components in the drywell. The stub tube was plugged, thus directing the plume back to the equipment sump and isolating the condensation of the steam to the drywell floor drain (unidentified leakage). Some leakage was still observed from the plugged stub tub Multiple drywell entries were made to determine the source of the leakage using hand held pyrometers and stethoscopes. The source of the leakage was determined to be the 833F025A and the 833F026A manual vent valves off of the recirculation suction gate valve 833F023. The leakage rate was estimated to be as high as 4 gpm. The F025A and F026A are in series on the vent lines prior to the line penetrating the stub tube. The vent /

drait, lines are 3/4 inch piping. The leaking valves could not be isolated from the reactor vessel since the recirculation valve (833-F023A) has a pressure relief opening on the reactor side of the double sided dis Efforts were made by plant staff maintenance personnel to open and reclose the up) stream (F026A vent could be valve (F025A)

mechanically suchand isolated thatrepaire the downstream vent valve The upstream valve could not be made to provide sufficient isolation to enable disc replace-ment on the downstream valve. The licensee initiated Material Nonconfor-mance Report 182-88 on September 9,1988, for the leaking valves by the seat to be repaired. MNCR 182-88 was dispositioned as "interim" to allow the seat of the valves to be temporarily sealed using on-liqe sealant injection technique by Furmanite. Thesealantinjectionwasmadethrough a nozzle inserted into the valve body (F026A) using a threaded connectio The valve body is ASME SA-182 F316 material which is compatible with the injection nozzle. The injection nozzle location on the valve was evalu-ated by NPE to verify that allowable valve body stress limits were not exceeded due to the hole location. The installation of the injection nozzle was accomplished such that the nozzle becomes a part of the valve pressure retaining boundary. The Safety Evaluation Report provided with the disposition of MNCR 182-88 was reviewed by the inspector, Attempts to seal the leakage through the valve by on-line sealant injection failed due to the high leakage rates. SERI reevaluated other means to isolate the leakage rates through the valves. Interim disposition #2 was provided to allow for the cutting and capping of the vent line (3/4" HCD-73) in order to stop the lekkage past the valves. The capping of the line consisted of a 1 1/4 inch socket weld coupling with a 3/4 inch threaded piae cap welded in one end and 3/4 inch socket weld pipe cap welded in the ot1er end. The threaded cap was inserted on the piping to the valves and the socket weld cap on the piping to the drain hub. The threaded cap was seal welded in order to reduce the possibility of leakage. MNCR 182-88 interim disposi-tion #2 also allowed for the installation of a Thaxton plug to the drain hub serving the vent lines since the required pipe cap could not be replaced on the hub due to damage to the threads. The actual work was controlled under WO MS4148 work instruction and inspection records. The inspector reviewed the 10 CFR 50.59 evaluation by the licensee for cutting and capping off the vent line. The capping of this line is considered temporary to permit continued plant operation until refueling outage 3 wherein a permanent repair will be made. Im31ementation and evaluat on of a permanent repair of the 3/4 inch HCO-731 ne prior to startup from RF03 will be identified as Inspector Followup Item 416/88-19-0 _ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

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At 8:15 a.m., on September 11, 1988, a Reactor Water Cleanup (RWCU) system isolation occurred as operators were shif ting from post-pump mode to '

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pre pump mode. The plant was in Operational Condition 3, Hot Shutdown, i

with reactor pressure at approximately 25 psig. In the post-pump mode i reactor water is cooled via the RWCU heat exchangers prior to entering the pumps to extend pump seal life. The "pre pump mode is where reactor water is drawn directly to the pumps prior to entering the heat exchanger The operators had secured tie RWCU pumps and were swapping from post pump mode to pre pump mode of operation per Procedure 101- directs the transfer of post-pump to pre pump at 25 psig reactor i pressure. The hi t h differential flow timer timed out causing closure of i all groups eight < solation valves on a leak detection differential flow -

signal. The RWCU system flow fluctuations were induced by having both '

pumps secured and having air trapped in the flow transmitters causing

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erroneous flow information. RWCU was restored to pre pumn mode of opera-1 tion. Similar events occurred on May 17, 1985, as dacribed in LER

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85-019-00 and on January 12, 1988, as described in LER 88-04-01. As part  ;

, of Supplemental Corrective actions associated with LER 88-04-01 the plant  ;

shutdown procedure, plant startus procedure, and the System Operating

! Instruction were changed to provide adequate margin in the pump suction a

pressure for RWCU pump shutdown without causing a high differential flow system isolation. All 101s were reviewed by the licensee to determine if ,

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the potential for creating low pump suction head exists in other opera-tions of the RWCU system. All licensed operations personnel were made  !

] aware of the significance of those changes and the potential for RWCU

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system isolations. In LER 88-04-01 supplemental corrective action stated ';

l that SERI determined the prudent design change to make is to install a i keylock bypass switch that bypasses the delta flow isolation signal. The l l licensee intends to use the keylock bypass switch during anticipated RWCU i i system cperati.1g transients (RWCU system startup, shutdown, etc.) which [

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could result in spurious isolation This design change is currently ;

scheduled for implementation during the third refueling outage. Instal- -

3 lation of a keylock bypass switch for delta flow will be tracked as

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Inspector Followup Item 416/88-19-03.

i On September 15, 1988, Division 2 Diesel Generator was scheduled for I

routine maintenance. It was observed by a SERI staff worker that water  !

! was leaking from the lef t bank intercooler. During investigation  ;

l activities under Maintenance Work Crder M84200 it was discovered that an  ;

internal diffuser in the lef t bank intercooler adapter had broken loose  ;

i from the adapter shell and dropped down onto the tube bundle in the L

, intercooler. Over a period of time the normal engine vibration caused l this piece to cut into the aluminum fins and brass tubes, apparently v

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causing the leakage. The internal diffuser plate is used to direct air  !

from the turbocharger into the intercooler and provide for even air  :

distribution over the intercooler tubes. Material Nonconformance Report i i 184-88 was written to document this nonconformance. Disposition instruc-i tions to MNCR 184-88 was provided by NPE and mechanical maintenance to j repair the intercooler adapter broken diffuser. Interim disposition f j number 1 for the repair process added additional connection surface for the diffuser plate to the adapter shell and reduced the amount of free j

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standing plate that would be susceptible to movemen The diffuser plate in the right bank intercooler adapter is of a completely different confi-guration from the left bank. Because of the angle at which the turbo-charger provides intake air to the right bank intercooler, the diffuser plate is configured to a much more complex geometry adding substantial rigidity. The diffuser plate creates a crucible configuration that runs the full length and width of the adapter shell. All ends and the majority of the top edge of the plates are attached to the adapter shell by welding. The configuration of the diffuser plate for the left bank intercooler adapter allowed sufficient movement during air flow to initiate a fatigue environment that resulted in failure. After partial failure of the diffuser platt, contact was made with the intercooler tubes causing fretting between the plate and tube surfaces. To preclude the possibility of similar occurrences the licensee inspected the Division 1 diesel generator left bank intercooler inlet adapter per MNCR 0184-88. On September 20, 1988, results of the licensee inspection showed that cracks existed in the vane area where the stiffener is attached to vane with weld No damage was observed to the intercooler. Crack repairs were performed in accordance with MNCR 182-88 instructions, Division 1 diesel was returned back to service on September 21, 1988. Upon discovery of cracks associated with Division 1 diesel generator inlet adapter, the licensee made a four hour courtesy call to the NRC duty office Prf.sently the licensee is evaluating the defects against 10 CFR Part 21 criteria. Determination on Part 21 of D/G cracks associates with the inlet adapter guide vane will be identified as Inspector Followup Item 416/88-19-0 . Inspector followup and Unresolved Items (92701)

(Closed) Inspector Followup Item 416/88-07-01 Correct discrepancies noted during walkdown of the Combustible Gas Control System. The licensee issued TCN-6 to Procedure 50I 04-1-01-E611 Revision 20. TCN-6 corrected valve numbers and instrument number for valve FX026 and FX027. P05-N001 was also corrected through TCN-6. The inspector verified TCN-6 corrected the discrepancies and valve E61-F031 was properly labele (Closed) Inspector Followup Item 416/88-12-01. The addition of identi-fying numbers to instruments. The inspector verified that the labeling was corrected on control panels 1H13-P669 anti 1H13-P67J for instruments 017-RITS-K621A, K609A, K617A, K618A, K6210, K609C, K617C and KC18 (Closed) Inspector Followup Item 416/88-17-02. Investigation of recir-culation pump "A" motor bearing oil level alarm. On August 15, 1988, when the reactor scram oil levels were checked in both A and 8 pump motors per NWO 83459, approximately 1.5 gallons of oil was added to the upper motor and the alarm cleare _ _ _ _ ____ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _

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i Exit Interview (30703) i

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The inspection scope and findings were summarized on September 23, 1988,

with those persons indicated in paragraph 1 abov The licensee did not L identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. The licensee had no comment on the i following inspection findings (Paragraph 7): i l

416/88-19-01, Violation, Failure to follow procedure for protective l taggin i 416/88-19-02, Inspector Followu) Item, Implementation and evaluation of a permanent repair of the vent /crain valves prior to startup from RF0 !

416/88-19-03, Inspector Followup Item, Installation of a keylock bypass ,

switch for delta flow on RWCU Syste !

416/88-19-04, Inspector Followup Item, Determination of a 10 CFR Part 21 report for adapter vane crack on Division 1 and 2 Diesel Generato l j 1 Acronyms and Initialisms ATWS -

Anticipated Transient Without Scram ASME -

American Society of Mechanical Engineering i BOP -

Balance Of Plant CR0 -

Control Rod Drive 0/G -

Diesel Generator 0/W -

Drywell ECCS -

Emergency Core Cooling System ESF -

Engineered Safety Feature FT -

Feet GETARS -

General Electric Transient Analysis Recording System GPM -

Gallon Per Minute 101 -

Integrated Operating Instruction LC0 -

Limiting Condition of Operation LER -

Licensee Event Report MNCR -

Material Nonconformance Report MSIV -

Main Steam Isolation Valve MWO -

Maintenance Work Order NPE -

Nuclear Plant Engineering ONEP -

Off Normal Event Procedure PA -

Public Address RPS -

Reactor Protection System RWCU -

Reactor Water Cleanup RWP -

Radiation Work Permit SER -

Safety Evaluation Report SERI -

System Energy Resource Incorporated SOE -

Sequence Of Events 501 -

System Operating Instruction TCN -

Test Change Notice WI&IR -

Work Instruction and Inspection Report