IR 05000416/1987018

Insp Rept 50-416/87-18 on 870718-0821.Violations Noted. Major Areas Inspected:Licensee Action on Previous Enforcement Matters,Operational Safety Verification,Maint Observation,Surveillance Observation & ESF Sys Walkdown

ML20237H175
Person / Time
Site:	Grand Gulf
Issue date:	08/28/1987
From:	Butcher R, Dance H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20237H149	List: IR 05000416/1987018 ML20237H146 ML20237H156
References
50-416-87-18, NUDOCS 8709030147
Download: ML20237H175 (16)

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pe af o UNITED STATES f vq'o NUCLEAR REGULATORY COMMISSION

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',' n REGloN 11 101 MARIETTA STREET, $ . j

• I 'c ATLANTA, GEORGIA 30323

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Report No.: 50-416/87-18 Licensee: System Energy Resources, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Nuclear Station Inspection Con uc d: July 18 through August 21, 1987 Inspe tors:- / C< m < 17 ~7

- . Butcher, Senior Rest' dent Inspector Date Signed

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, e J. L. Mathis, Resident Itrspector Date Signed Accompanying Inspector: L. P. Modenos, Project Engineer (July 20-24, 1987)

Approved by: C 0h-ff. C. Dance', Section Chief V

Date Signed I 77 Division 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, Operational Safety Verification, Maintenance Observation, Surveillance Observation., ESF System Walkdown, Reportable Occurrences, Operating Reactor Events, Inspector Followup and Unresolved Items, and Preparation for Refuelin Results: One violation was identified - failure to report a major loss of the alert notification syste PDR B70828 G ADOCK 05000416 pop

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' Licensee Employees Contacted i J. E. Cross, GGNS Site Director

• C. R. Hutchinson, GGNS General Manager i

R. F. Rogers, Manager, Special 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. McAnulty, 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. Houlder, Operations Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent S. M. Feith, Director, QA J. C. Cesare, Manager, Licensing

• R. C. Fron, Manager, Corporate Emergency Preparedness

• F. W. Titus, Director, Nuclear Plant Engineering

• S. F. Tanner, Manager, Quality Services

• R. H. Halbach, Administrative Assistant

• J. Czaika, SMEPA, Nuclear Specialist Other licensee employees contacted included technicians, operators, security force members, and of fice personne * Attended exit interview Exit Interview (30703)

The inspection scope and findings were summarized on August 21, 1987, with those persons indicated in paragraph I abov 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:

416/87-18-01, Inspector Followup Ite Discrepancies noted during walkdown of the control room HVAC system (paragraph 7) '

416/87-18-02, Unresolved Ite Reactor Water cleanup isolation valve power, annunciation and FSAR discrepancie (paragraph 9)

416/87-18-03, Inspector Followup Ite Discrepancies in post-trip analysis for scram No. 45. (paragraph 9)

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416/87-18-04, Inspector Followup Ite Dispositioning of dropped fuel bundles. (paragraph 11)

416/87-18-05, Violatio Failure to report a major loss of the alert notification syste (paragraph 9)

416/87-18-06, Inspector Followup Ite Remote shutdown panel room door discrepancies. (paragraph 10) Licensee Action on Previous Enforcement Matters (92702)

(Closed) Violation 416/86-32-04. The licensee gave General Electric refueling supervisors training on pertinent plant administrative procedures and administrative procedure 01-S-07-1, Control of Work on Plant Equipment and Facilities, was revised to require impact statements for specified conditions be given an independent review. No further action is require (Closed) Violation 416/84-45-02, Inadequate Fire Brigade Trainin Inspection report 86-24 reviewed the actions taken by the licensee to resolve the violatio The inspector acknowledged that the Fire Protection Training Program, Procedure No.10-S-03-7, was issued which defines and implements the fire brigade training program. However, the inspector identified that the new tracking program did not include a review of quarterly classroom training and fire brigade drills. In addition, the Radwaste personnel included in the list of qualified personnel for emergency worker and fire brigade duties were on different shifts from the Operations group. The licensee resolved these concerns by creating a quarterly meeting which is tracked by a computer program. This program lists all personnel that are required to attend the next quarterly meeting. A listing of all unqualified personnel is printed monthly and sent to the Shift Superintendent / Supervisors. The Radwaste personnel are no longer routinely assigned to the Fire Brigade although they are assigned to Emergency Wor However, the Radwaste personnel are still being trained for Fire Brigade duty and may be drawn into the pool if needed, after verification of their qualification The inspector ,

reviewed the tracking system, Fire Protection Training Procedure, the quarterly printout and the monthly memo to Shift Superintendents. The inspector concluded that the corrective actions taken are adequate and that the violation is close . Operational Safety, Radiological Protection and Physical Security 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 operations. Daily discussions were held with plant management and various members of the plant operating staf ____ _ _ ___ _ _____ ________a

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The inspectors made frequent visits to the control room such that it was visited at least ' daily when an inspector was onsit Observations included instrument readings, setpoints and recordings, status of <

operating systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation, 4 temporary alterations in effect, daily journals and data sheet entries, 1 control room manning, and access control This inspection activity

)j included numerous informal . discussions with operators and their supervisor '

Weekly, when the inspectors were onsite, selected Engineered Safety _ l Feature (ESF) systems were confirmed operable. 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 visite Observations included safety related tagout verifications, shif t turnover, sampling program, housekeeping and general plant conditions, fire protection equipment, control of activities in progress, problem identification systems, and containment isolation. At least monthly, the licensee's onsite emergency response facilities were toured to determine facility readines Monthly, the inspectors reviewed at least one Radiation Work Permit (RWP),

observed health physics management involvement and awareness of significant plant activities, and observed plant radiation controls. At least quarterly the inspectors reviewed the licensee's program to limit f.ersonnel radiation exposure As low As Reasonably Achievable (ALARA).

Monthly, the inspectors verified licensee compliance with physical security manning and access control requirements.- At least quarterly j the inspectors verified the adequacy of physical security detection and assessment aid No violations or deviations were identifie . Maintenance Observation (62703)

During the report period, the inspectors observed portions of ' the 1 maintenance activities listed belo The observations included a 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 MWO I73610, Low sample flow on drywell hydrogen analyzer A. (LC0 87-665).

MWO 173499, Recirculation pump motor A/B temperature No. 1 alarming.

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.MWO 173917, 173919 and 173920. Install thermocouple on penetrations 83, 87, 88, and both reactor water clearup pumps and monitor temperature No violations or deviations were identifie . Surveillance Observation _(61726)

The inspectors observed the performance of portions of the surveillance 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 surveillance, removal'from service and return to service of the system or coniponents affected, and review of the data for. acceptability based upon the acceptance criteri ME-1M23-V-0001, Revision 2 Containment and Airlock Seal Leak Tes IC-1E61-M-1004, Revision 26, TCN 12. Containment and Drywell Hydrogen Analyzer (PAM) Ca MWO IND601, Control Room A/C Unit Heat Exchanger Flo IC-1C34-M-0001, Revision 24. Reactor Vessel Water Level High (Level 8)

MT/RFPT Tri EL-1L11-W-0001, Revision 24, 125 Volt Battery Bank Pilot Cel RE-1C51-0-0001, Revision 26, Local Power Range Monitoring Calibratio No violations or deviations were identifie . Engineered Safety Features System Walkdown (71710)

A complete walkdown was conducted on the accessible portions of the Control Room Heating, Ventilation and Air Conditioning (HVAC) System. 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 and evidence of rodents, and system free from other degrading condition The system was functional and would have performed its safety function. The following minor discrepancies were identified:

Piping and Instrumentation Diagram (P&ID) M-0049, Revision 19 shows the following manual valves but they do not have a C indicating normally close Z51-F024A, F0248, F029A, F0298, F032A, F032B, i F038A, F038B, F039A, F039B, F040A, and F040 Service valve Z51-F024A and F0248 appear to be mislocated on drawing M-0049, Revision 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _

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l Water was found covering the floor in the N040B chlorine detector room on the 189 foot elevation of the Control Building. MWO 73683 was initiated to correct the proble System Operating Instruction 04-S-01-Z51-1, Revision 23, attachment III, page 2 lists component F0128. The number should be F01 l Followup on the correction of the above discrepancies will be tracked by Inspector Followup Item 416/87-18-0 No violations or deviations were identifie . Reportable Occurrences (90712 & 92700)

The below listed event reports were reviewed to determine if the information provided met the NRC reporting requirement 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 asteris The event reports were reviewed using the guidance of the general policy and procedure for NRC enforcement actions, regarding licensee identified violation The following License Event Reports (LERs) are close LER N Event Date Event

• 87-008 May 11, 1987 Residual Heat Removal A Pump Incorrectly Tested Due to Personnel Erro No violations or deviations were identifie . Operating Reactor Events (93702)

The inspectors reviewed activities associated with the below listed reactor event The review included determination of cause, safety significance, performance of personnel and systems, and. corrective actio The inspectors examined instrument recordings, computer printouts, operations journal entries, scram reports and had discussions with operations, maintenance and engineering support personnel as appropriat On July 29, 1987, Material Nonconformance Report (MNCR) 0231-87 was written to document the licensee's discovery that at containment penetration 87, outboard containment isolation valve G33-F004 and inboard containment isolation valve G33-F252 were both powered from the Division 1 power supply. The Final Safety Analysis Report (FSAR) paragraph 6.2.4. states that for power operated valves used in series, no single event can interrupt motive power to both closure devices FSAR paragraph 7.3.1.1.2.2 states that power for the operation of two redundant valves in

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a line is supplied from separate ESF buses. Incident Report (IR) 87-7-14 was initiated by the Shift Superintendent to document the potential problem and Limiting Condition for Operation (LCO)87-700 initiated the action statement of TS 3.6.4. The licensee secured the Reactor Water Cleanup (RWCU) system, tagged closed and removed power to G33-F252. The inspector observed that when the breaker supplying power to isolation valve G33-F252 was tagged opened the control room annunciator P601-19A-H3, RX OTBD ISOL SYS 00SVC, annunciated and the outboard NSSSS status light on panel P601-198, OTBD MOV OVERLD/PWRLOSS, illuminated. The noted alarms would indicate an outboard isolation valve problem but valve G33-F252 is located inside the drywell and is in reality an inboard isolation valv The reason for this human factors discrepancy is that although convention (See FSAR figure 7.1-4) calls for outboard isolation valves to be powered from Division 1 power, valve G33-F252 is an exception in that it is powered from Division 1 power. The' licensee had been operating in what is called the Post Pump Mode of RWCU operatio An alternate RWCU operational mode, called Pre-Pump Mode, uses a different flow path and eliminates the noted problem but allows the reactor coolant to go directly to the RWCU pumps without going thru the RWCU heat exchangers which imposes a higher temperature on RWCU pump components. It is noted that the FSAR description of RWCU system operation only addresses the Pre-Pump Mode which the licensee has not been using for operation. A potential problem with adequate cooling to containment penetrations due to fouling problems from the Plant Service Water (PSW) system was previously addressed in Inspection Report 86-32 and followed as IFI 86-32-02. This item was discussed in Inspection Report 86-39 where it was noted only penetration 83 would experience high temperature fluid flow. Based on temperature measurements this IFI was closed in Report 87-01. With the revised flow path of RWCU, penetrations 87 and 88 will now also experience high temperature fluid flow. As discussed in FSAR paragraph 6 2.1.1.10, a temperature of less than 200 F anywhere in the concrete adjacent to the penetration is considered acceptable. On July 30, 1987, the licensee installed temperature sensors on penetrations 87 and 88 in order to monitor penetration temperatures when the RWCU system is put in operation in the Pre-Pump Mode. The licensee has initiated a Corrective Action Plan for the issues generated from this discovery. The immediate actions taken were:

Monitor RWCU component and system performance for possible deterioration in the Pre-Pump Mode of operatio Monitor temperatures of containment penetrations affected by the Pre-Pump mode of operatio Initiate a root cause assessment since this discrepancy has existed since 1979 time fram Notified other BWRs of RWCU isolation issu Evaluate for part 21 deportability.

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The licensee documented the discrepancy of both the inboard and outboard containment isolation valves being powered from Division 1 power on Incident Report (IR) 87.-7-14 dated July 29, 1987, and took prompt and adequate corrective action. The licensee's final resolution of operation without adequate containment isolation will be tracked as Unresolved Item 416/87-18-0 On July 23,1987 at 7:10 p.m. during the performance of the Division 2 Diesel Generator (DG) monthly surveillance (06-0P-IP75-M-0002-3), the DG tripped 58 minutes into the run. The procedure requires a one hour ru Local alarms indicated the trip was due to high vibration. Instruments-tion and Controls -(I&C) technicians were taking vibration readings at the time of the trip. The portable vibration meter magnetic probe used for vibration measurement was placed too close to the installed vibration probe on the right side of the turbocharger creating a false high vibration signal . The Division 2 DG was restarted and the surveillance successfully complete The licensee considers this a non-valid failur On. August 3, 1987 at 3:31 p.m. the operators noticed that main steam line safety relief valve logic A trip unit B21-PIS-N668A tripped for no apparent reaso This is one half the signal required.for a Division 1 initiation of safety relief valve B21-F051D which has a low-low set function to open at 1033 + 15 psig and reset at 926 + 15 psig. Reactor pressure at high power levels is approximately 1030 psig. Similar events have occurred several times recently, on July 17, July 19, and July 27, 1987. On January 23, 1987, six safety relief valves inadvertently lif ted during the performance of the SRV High Pressure Trip / Low-Low Set Relief monthly functional surveillanc This event was determined as not reportable by the licensee but is discussed in a letter to the NRC dated March 2, 1987 (AECM-87/0044). The licensee is - considering two modifications to reduce the potential for SRV challenges. Design Chang Requests (DCRs) 87/037 and 87/038 provide arc suppressors and individual conducters to terminals from the common buss and are scheduled for incorporation during refueling outage number 2 scheduled to begin on November 6, 1987. The event of January 23, 1987 was previously discussed in Report 87-01 and is being tracked as IFI 416/87-01-06. These-latest events will be tracked with this IF On August 5,1987 at 8:27 a.m. the Reactor Protection System (RPS) motor generator A output breaker (C71-S003C) tripped giving a reactor half scram. Annunciator 7A-A2 on panel 1H13-P680, RX SCRAM TRIP, sealed i The operators reset breaker C71-S003C which is located in the motor generator A room. The room temperature was measured at 114 F and it was felt that the high temperature was the cause of the breaker tri Operations reviewed their tagouts and determined that dampers controlling cooling air to the room had been tagged closed. The tags were lifted and cooling air to the room restore __ . _ _ . ._ __ ______ -_____-___-_.-_ _ ___-_ ___ - -__

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On August 6, 1987 at 6:25 a.m'. while operating at 100 % reactor power, the reactor scramme A load. reject signal caused a turbine control valve fast closure which initiated a reactor scram. Reactor pressure reached 1107 psig and six safety reitef valves opened. No Emergency Core Cooling Systems (ECCS) actuated. The load reject was the result of the opening of the main generator output breakers (J5228 and J5232). A lockout relay associated with both breakers was found tripped. The lockout relay is out of the circuit when breaker J5230 is closed which is the normal operation lineup. Due to a buildup of moisture and corrosion in a terminal cabinet the lockout relay was actuated which caused breakers J5228 and J5232 to open. After coordination with MP&L the lockout relay was. isolated from the breaker circuitr The cabinets are located in n environmentally controlled building but the moisture buildup occurred due to openings for the cable pathway and the cabinet doors were kept closed. The immediate fix was to open the cabinet doch to allow ventilation and the cleaning of the terminals. The inspectors reviewed the Post-Trip Analysis for the scram (Scram No. 45). The analysis was adequate but several areas were incomplete and reflects pour review by manegement. The areas in question were:

Paragraph 1 states. to obtain the listed data and if unavailable, explain. There were several pieces of da ta unavailable but no explanation was give Paragraph 2 has a , spa $e for entering the parameter causing t~ n e tri It appears computer pair.t numbers were listed wMch would not be understood by most: people reviewing the analysi Paragraph 6 states that if one or more SRVs lifted, list the valve number (s) and time of opening. If SRVs lifted more than once list combinations and times. Include if low-low set functioned properl The SRVs that opened were listed but the cycling of the B21-F0510 was not noted. No times were given for SRVs lifting. No comment was made regarding if low-low set functioned properl Paragraph 10 asks what is the probable cause of the reactor tri The cause listed was a load reject signal caused by a turbine control valve fast closure. The ROOT CAUSE was not give Paragraph 13 classifies the event into several categories where the cause of the trip is known, unknown, safety equipment malfunctioned, et It is not clear from paragraph 10 above that the root cause must be known to classify the event and determine if it is acceptable to restart the reacto The inspectors w1?1 review the licensee's actions on the noted comment This action will be tracked as Inspector Follopp Item 416/87-18-0 o

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l On August 6, 1087 while the plant was in hot shutdown the licensee identified that the ductwork between the high energy check.darpers and the building penetrations had nok been analyzed to demonstrate it is capable of withstanding the pressure dif ferentials to which it might be subjecte The damper installations in c;uestion were:

Safeguard switchgear and battery room HVAC system (Q1Z77F005A, F005B, Q2Z77F005A, F0.05B, Q1Z77F006A, F006B, Q2Z77F006A and F006B).

ControlRoomairintake(QSf51F0t).,

RHR A and B room supply (11T42F024, F031, F032 and FU33).

RHR A and B room exhaust (Q1T42F025 and F026).

RCIC room supply (Q1T42F023).

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, i RCIC room exhaust (Q1T42F022).

RWCU pump room supply (Q1T42F028 and F029).

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RWCU pump room exhaust (Q1T42F027 and F939).

FSAR paragraph 3.3.2 states that tornado resistant seismic category 1 structures are analyzed for a maximum design pressure drop of 3 psi with a maximum rated change of 2 psi /second. The containment and control building are not vented structures and the exposed exterior walls of these structures are designed for the 3 psig pressure drop. The licensee's corrective action was to remain in hot standby while modifying the ductwork to dampers Q1Z77F006A, F006B, and 02Z77F005B to ensure they were capable of withstanding a 3 psi differential pressure. Calculations indicated the other exterior wall penetrations were adequate. This is a licensee identified deficiency and wac reported in a timely manner and was immediately corrected prior to returning to power. The inspector evaluated the significance of the event and licensee actions taken and concluded that the five criteria listed in Section 5.1 of 10 CFR 2, i Appendix C, General Statement of Policy and Procedure for NRC Enforcement !

Actions, have been met, thus no notice of violation will be issue !

On August 3, 1987 during a monthly offsite siren test the Claiborne County sirens did not actuate when the initiate button was pushe Claiborne j County, Ms. , has 30 sirens and Tensas Parish, La. , has 13 sirens. The !

Tensas Parish siren test was successful. The licensee had a repair crew I investigate the Claiborne County problem within two hours and they could l not identify a specific proble The sirens were tested again I (unofficially since no notice of siren test could be announced so soon) I and the sirens could be heard going off but no record of how many sirens functioned was obtained. It was August 10, 1987 before a notice could be published and a second official offsite siren test could be conducte The second test was successful with 23 out of 30 Claiborne County sirens

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i operating properly. It is noted that the success criteria of 70 % or

% grnter is based on Claiborne County and Tensas Parish sirens combined.

8- Therefore, if all 13 Tensas Parish sirens failed to alarm, the test could'

b stil l. be successful based on all 30 Claiborne County s'i ren s being

$" successful. On August 17, 1987 technicians found a potentially bad connector to the antenna and replaced the - connecto NUREG-0654, Revision 1, Appendix 3, paragraph C.3.h states as guidance a growl test 4 (or equipment) should be conducted quarterly and when preventive

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maintenance is performe The GGNS Emergency Plan, Revision 15, 3 paragraph 8.6, Maintenance and Inventory of Emergency Equipment and Supplies, states all emergency equipment / instruments shall be inspected, f inventoried and operationally tested on a quarterly basis and after.each

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9 % o y rability in June, 1987 and 95% operability in July 198 It appears they have met the recommendations in NUREG-0654 and the requirements specified in their Emergency Plan. When the sirens failed to f( initiate on August 3, 1987 the Claiborne County Civ11 Defense notified the licensee's corpcrate office of the problem. An agreement letter between the licensee and the Civil Defense personnel calls for the licensee to be notified of test result The licensee's corporate personnel were not aware of the deportability requirement During subsequent discussions regarding the event, the que'stion of deportability was raised and the WPC was notified on August 14, 1987. This was documented on Incident Report 87-8-7. Administrative Procedure (AP) 01-S-06-5, Revision 15, Incident Reports / Reportable Events, attachment III, paragraph I.7 which references attachment-V, page 18, requires immediate notification (within one hour of occurrence) of any event where the siren portion of the offsite alert

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notification system has less than 70% siren operability. 10 CFR 50.72 (b)(1)(v) requires the NRC be notified as soon as practical and in all cases within one hour of any event that results in a major loss of '

emergency assessment capability, offsite response capability, or communications capability. The failure to report the loss of the offsite siren system during the August 3, 1987 test is a violation of this j requirement. This is identified as violation 416/87-18-05. Also, part of' j the alert notification system per the licensee's Emergency Plan, paragraph 7.5.4, is the institutionally located tone activated receivers. The tone activated receivers also failed to actuate Juring the August 3, test. The licensee's Ap 01-S-06-5 does not address the loss of tone activated I i

receivers. Furthermore, the licensee does not have a proceaure covering the conduct of testing the alert notification system nor what actions to take at what threshold. The inspector discussed the problems noted with the licensee and the licensee has developed a long term and short term corrective action pla This plan includes procedural, training and equipment change The licensee committed to submit these commitments formally in their reply to the violatio >

On August 18, 1987, at 00:50 a.m. while operating at 100% reactor power, radial well pumps J and K tripped. The plant normally operates with seven radial well pumps in operatio The wells and associated pumps are as follows:

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i Well No. 1 Radial Well Pumps A&B Well No. 3 Radial Well Pumps C&D i (Only pump C was operating) i Well No. 5 Radial Well Pumps E&F Well No. 4 Radial Well Pumps J&K Plant chillers A, B and C tripped due to low Plant Service Water (PSW)

flow. Radial well pump D was started fro- the control room and Chillers A

& B were restarted but at 1:10 a.m. Pump C tripped as did plant chiller The auxiliary building steam tunnel temperature was observed to be 105 F and increasing so power was reduced to approximately 70% to reduce heat loads. Investigation revealed no level indication for well number 4 due to failure of tubing for the bubbler level system giving a false low well water level signal and a trip signal to the pumps. Operations initiated the following actions to reduce PSW loads:

The Fuel Pool Cooling and Cleanup (FPCCU) system was taken off Component Cooling Water (CCW).

Control room air conditioning and the A ESF switchgear room coolers were switched to the Standby Service Water (SSW) system. The SSW pump A was started to supply this equipmen Circulating water blow down flow was reduced to stabilize cooling tower basin leve The J and K radial well pumps were restored after troubleshooting and repairing faulty control components. Well number 3 normally operates with only one pump due to problems maintaining adequate well water level and is scheduled to be refurbished during the upcoming refueling outag Incident Report 87-8-8 was written to document this event for licensee evaluatio . Inspector Followup and Unresolved Items (92701)

(Closed) Inspector Followup Item 416/84-15-01, Review of report on abnormal snubber behavior during valve stroke testing. The inspector reviewed the report dated August 10, 1984, MWO-M44119 and Pacific Scientific evaluation report (PQIPN-R0868-84) and verified that the six snubbers in question were evaluated, passed functional testing and/or refurbishe (Closed) Inspector Followup Item 416/85-16-04, Inadequate number of eight (8) hour emergency lighting units. The inspector reviewed the Design Change Package (DCP) 85/3100 to verify implementation of the eight hour emergency lighting for access to and operation of alternate shutdown panels in the Control Building, Auxiliary Building and Diesel Generator Buildin DCP 85/3100 was also reviewed by the resident inspectors as

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Licensee Condition (LC) 2.C.(22) and was addressed as open item 86-32-06 in Inspection Report 86-32. This item was subsequently reviewed and installations were verified by a walkdown and 86-32-06 was closed in Inspection Report 86-3 The design change implementation records were verified to be accurate and with the walkdown done by the resident inspectors, satisfies the disposition of this inspector followup ite (Closed) Inspector Followup Item 416/84-15-03, Review the licensee's evaluation of the vibration test program. As a result of two cracks found in the 3 inch diameter pipe of the RHR B loop, the plant issued LER 84-24 describing the event. Evaluation of the RHR system is described in GGNP Report No. 84-M-005, August 198 The report summarizes the RHR system evaluation for the abnormal vibrations experienced during the plant testing at low power. Several corrective actions have been taken to preclude this problem:

High point vents have been added to pump suction and discharge piping to facilitate the removal of gas void Startup procedures (RHR Procedure No. 04-1-01-E12-1 and LPCI/RHR Procedure No. 06-08-1E12-Q-0023), have been revised to:

Preclude throttling of certain valves (F003 and F048) in certain modes to minimize the flow induced vibration Reduce flow rate from 2000 gpm to 1200 gpm before shutting off the pump to minimize check valve nois Pressurize the suction pipin The inspector reviewed the above procedures and vibration test report and ,

considers the item close ;

(Closed) Inspector Followup Item 416/85-29-02, Include a revised curve of Power-Flow performance with a discussion of the reasons for and significance of differences from FSAR Figure 14.2-4 in the Startup Repor See IFI 416/85-29-03 below.

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(Closed) Inspector Followup Item 416/85-29-03, Concurrent with the i submission of the Startup Test Report, submit proposed changes to FSAR )

Figure 14.2-4 and TS Figure B 3/4.2.3-1 for review by NRC/NRR. The two related items referenced above have been reviewed by the Licensee and General Electric (GE) and they concluded that the Power / Flow (P/F) Map is being used in the Control Room as an aid. It has no safety uses, and there is no need or requirement for its use. The purpose for the maps in the UFSAR is strictly illustrative in nature, to provide pictorial presentation of typical Power / Flow profile. In order to clarify the use of these maps, a footnote was added to Figure 14.2-4. The footnote provides reference to the Startup Test Report which contains the Unit I l Power / Flow Map generated from the unit's Startup Test Program (See

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licensee letter AECM-86/0066). A footnote was also added to the Power /

Flow map appearing in Chapter 4 (USFSAR Figure 4.4-5) to clarify the figure's purpose, i.e. to present the typical Power / Flow relationship or

" operating regions" of a BWR/6. The inspector reviewed the letters from GE, MPGE-86/163 and NLSIM-86/0779 and after conferring with the Region II management concluded that the two items are satisfactorily dispositioned as close (Closed) Inspector Followup Item 416/85-16-02, Review of licensee's actions to provide total enclosures for Appendix R safe shutdown cable raceway fire barriers. The licensee conducted a walkdown and identified ten (10) fire zones which the rae.eway/ support interfaces were found to be inadequately wrapped. The inspector reviewed the Material Nonconformance Report (MNCR)0267-85 and associated MWRs, F61972, F61973, F61974, F61975, and F61976 for complianc The modifications appeared to correct the discrepancies identified, therefore, this item is close (Closed) Inspector Followup Item 416/86-39-05. The licensee committed in a letter dated November 24, 1986 (AECM-86/0363) to perform an additional ,

test to demonstrate the ability to maintain the Remote Shutdown Panel (RSP) rooms at an effective temperature of 85 F or less for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with an ambient outdoor temperature of at least 95 The definition of an effective temperature of 85 F was defined in a letter dated September 13, 1982 from D. Tondi, Acting Branch Chief in the Division of Human Factors Safety to A. Schwencer, Branch Chief in the Division of Licensin This test is part of License Condition (LC)

2.C(33)(a). The licensee issued Technical Special Test Instruction (TSTI)

1Z77-87-001-1-S to conduct the noted test. Subsequent to the report period, on August 23, 1987, the licensee conducted TSTI 1Z777-87-001-1- The test results appear to satisfy the acceptance criteria in LC 2.C(33)(a) and this item is closed. However, the inspector noted the following discrepancies during the tes The TSTI, paragraph 7.1 refers to RSP rooms OC208 and OC208A. The door between the two RSP rooms has an identification tag identifying the OC208A room as OC20 For the RSP room airconditioner to function properly, the isolation door (0C223) between rooms OC208 and OC208A should be left in the open position since air is discharged into room OC208 and the suction is from room OC208A. The door does not have a warning label to leave it in the open position but does have a sign indicating the door is a TS fire door which might influence personnel to close the door. The inspector was told the door would automatically close in case of fir The resolution of these two discrepancies will be tracked as Inspector Followup Item 416/87-18-0 _ ._

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1 Preparation for Refueling (60705)

On July 31, 1987 at 3:15 p.m. while transferring a Reactor Assemblies (RA)

container from the turbine building railroad bay up to the 166 foot elevation of the turbine building the RA was dropped approximately three feet. The RA is approximately 17'3" x 2' 8" x 2' 9", weighs approximately 2800 pounds and contains two fuel bundles. A hoist with a 4 legged sling was used to lif t the RA from the railroad bay (elevation 133 feet) up to the 166 foot elevation where the fuel was then lowered onto a four wheel cart for movement to the refuel floor. After removing the four hooks from the RA and attempting to raise the hoist and sling up and out of the way, one hook caught on the RA cart and tipped the RA and cart over. The licensee then examined the RA and finding no visible or radiological evidence of damage, moved the RA to the refuel floor as previously planne The RA was opened and the two internal accelerometers were tripped. They trip at a value of 10 No visible damage to the fuel was noted and all radiological surveys showed no increased counts above normal backgroun Af ter talking to the fuel vendor, advanced Nuclear Fuel ( ANF), the licensee repackaged the dropped RA and is returning the two affected fuel assemblies to the vendor for further examination. The dispositioning of the affected fuel bundles will be Inspector Followup Item 416/87-18-04. Incident Report (IR) 87-7-15 was written to document this event. Although there was limited potential for ef fecting the health and safety of licensee personnel and/or the public from this event it is noted that other fuel movements could have the potential for much more severe consequences. It is therefore recommended that the licensee review fuel handling training and procedures to prevent recurrenc No violation or deviations were identified, gm " . ; F1

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