IR 05000255/1986023

Insp Rept 50-255/86-23 on 860723-0908.Violations Noted: Failure to Adequately Control Maint Activities in Accordance W/Procedures & Instructions & Failure to Adequately Test Following Maint

ML20214W209
Person / Time
Site:	Palisades
Issue date:	09/25/1986
From:	Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20214W188	List: IR 05000255/1986023 ML20214W185 ML20214W189
References
50-255-86-23, IEB-84-02, IEB-84-2, IEB-86-001, IEB-86-002, IEB-86-1, IEB-86-2, NUDOCS 8610030016
Download: ML20214W209 (15)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-255/86023(DRP)

Docket No. 50-255 License No. DPR-20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name: Palisades Nuclear Generating Plant Inspection At: Palisades Site, Covert, MI Inspection Conducted: July 23 through September 8,1986 Inspectors: E. R. Swanson C. D. Anderson Approved By: ef Reactor Projects Section 2A i d$I%

Date Inspection Summary Inspection on July 23 through September 8,1986 (Report No. 50-255/86023(DRP))

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Areas Inspected: Routine, unannounced inspection by resident inspectors of followup of previous inspection findings; operational safety; maintenance; surveillance; engineered safety features walkdown; reportable events; and Bulletin Results: Of the seven areas inspected one violation with three examples was identified for failure to adequately control maintenance activities in i accordance with procedures and instructions and one example of failure to adequately test following maintenance. Three unresolved items were identified which require further review for potential escalated enforcement actio The first concerns the inoperability of the containment air coolers due to inadequate service water flow and inadequate air flow. The second deals with the inadequate performance capability of the "B" low pressure safety injection pump. The third issue is over the inoperability of the component cooling water system as it relates to flow requirements of the shutdown cooling heat exchanger. Three open items were also identified to track completion of modifications to correct design errors and completion of corrective action .

8610030016 860926 5 PDR ADOCK 0500 O

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DETAILS 1. Persons Contacted Consumers Power Company (CPCo)

• J. F. Firlit, General Manager

• J. G. Lewis, Plant Technical Director

• R. D. Orosz, Engineering and Maintenance Manager W. L. Beckman, Radiological Services Manager C. E. Axtell, Health Physics Superintendent

• R. M. Rice, Plant Operations Manager

• R. A. Fenech, Plant Operations Superintendent

• H. M. Esch, Plant Administrative Manager S. C. Cote, Plant Property Protection Supervisor ,

J. R. Bradshaw, Property Protection Operations Supervisor

• K. E Osborne, Technical Engineer

• D. G Malone, Licensing Engineer

• R. A. Vincent, Plant Safety Engineering Administrator

• R. P. Margol, Quality Assurance Administrator

• T. J. Palmisano, Plant Projects Superintendent

* Denotes those present at the Management Interview.

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Other members of the Plant OperaLkns, Maintenance, Technical, and Chemistry Health Physics staffs, and sev eal members of the Contract Security Force, were also contacted briefl . Followup on Previous Inspection Findings (Closed) Open Item 255/85003-19: Administrative Procedure 2.01,

" Processing of Maintenance Orders," failed to specify the requirement of ANSI N18.7(1976) that maintenance and modifications are to be preplanned and performed in accordance with written procedures, instructions or drawings. Administrative Procedure 2.01 has been replaced with ,

Procedure 5.01, " Processing Work Requests / Work Orders." Administrative :

Procedure 5.01 includes the above requirement in Attachment 4 '

(Closed) Open Item 255/85015-01: During sampling of the safety injection tanks, the operators were continuing to drain past the Technical Specifica- i tion (TS) lower level limit thus entering a Limiting Condition for l Operation (LCO). The intentional entering of the LCO condition was not addressed in a reviewed and approved procedure. Technical Specification Surveillance Procedure MC-118, " Safeguards Boron Sample, Safety Injection Tanks," has been revised to address the intentional entering of the LC0 and limits the time of draining after the low level alarm is receive (Closed) Open Item 255/85009-02: "As found" valve stroke times are now required to be recorded. This requirement has been incorporated in recent revisions to Technical Specification Surveillance Procedures QO-2 I and QO-6. QO-5 and Q0-10 incorporate the requirement by the use of l temporary procedure changes which will become permanen . .

3. Operational Safety The inspectors observed control room activities, discussed these activities with plant operators, and reviewed various lons and other operations records throughout the inspection. Con rol room indicators and alarms, log sheets, turnover sheets, and iquipment status boards were routinely checked against operating requirement Pump and valve controls were verified to be proper for applicable plant conditions. On several occasions, the inspector observed shift turnover activities and shift briefing meeting Tours were conducted in the turbine and auxiliary buildings, and central alarm station to observe work activities and testing in progress and to observe plant equipment condition, cleanliness, fire safety, health physics and security measures, and adherence to procedural and regulatory requirements. A few fire extinguisher discrepancies were noted during routine tour These were discussed at the management intervie The inspectors made observations concerning radiological safety practices in the radiation controlled areas including: verification

! of proper posting; accuracy and currentness of area status sheets;

! verification of selected Radiation Work Permit (RWP) compliance; and implementation of proper personnel survey (frisking) and contamina-tion control (step-off pad) practices. Health Physics logs and dose records were routinely reviewe The inspectors observed physical security activities at various access control points, including proper personnel identification and search, and toured security barriers to verify maintenance of integrit Periodic observations were made of access control activities for vehicles and packages. Observations of activities in the Central Alarm Station were also conducce An ongoing review of all licensee corrective action program items at the Event Report level was performe During the use of the containment escape lock following preventive maintenance a licensee employee observed the outer door equalizing valve operating when the inner door was opened. The licensee made a 10 CFR 50.72 notification of this discovery on July 24, 1986, since it was considered a compromise of containment integrity that likely existed during periods of operation when containment integrity was required by Technical Specification 3.6.a. Subsequent investigation revealed an original design defect and that the as-built condition was accurately reflected on the drawings. The Palisades containment thirty inch escape airlock was manufactured and installed in accordance with W. J. Wooley Company drawings. A design feature of the lock was that a single operating shaft would open either door and the appropriate equalizing valve would equalize the air pressure across the door prior to openin It was found that the interlock

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caused the opposing door's equalizing valve to open resulting in an approximate one inch hole in containment whenever either the inner or outer door was opened. The problem existed since installation of the airloc The licensee completed a field machined modification to the roller cam which had caused the mis-operation of the opposing door valv Testing was done to verify the adequacy of the modification subsequent to installatio A verbal 10 CFR Part 21 report was made by the licensee to the NRC Region Administrator on August 8, 1986. The written followup report was submitted on August 13, 1986. This report outlined the design defect and listed other utilities which were likely affecte Technical Specification (TS) 3.6.1 which requires containment integrity when not in cold shutdown does not have an associated action statement if containment integrity is not met. Therefore TS 3.0.3 would be applied due to circumstances in excess of those specified for a limiting condition for operation and/or its associated action requirements. TS 3.0.3, which allows one hour to initiate action to shut down the plant, was likely never exceeded by the operating test or use of the airlock since an entry takes only a few minutes. The airlock interlock problem was identified by the licensee and promptly reported to the NRC upon discovery. When the vendor was notified of the situation and the apparent design error, he was reluctant to notify the NRC. The licensee subsequently provided the 10 CFR 21 notifications. There were no surveillance or operational tests that would have identified the design deficienc The design error was corrected by modification prior to resuming operatio LER 86-023 was submitted on August 22, 1986, providing additional details and the analysis of the situation. Both of the above mentioned reports (LER and Part 21) are considered close The TS violation is being considered for escalated enforcemen c. On July 25, 1986, while attempting to calibrate a low suction pressure transmitter on the P-8C auxiliary feedwater pump (AFP),

it was discovered that the loop polarity was incorrect due to lead reversal. This caused the output to indicate high at all times, placing the trip logic in a two of two instead of two of three for the AFP trip on low suction. It was determined that the last maintenance on this instrument was performed during the previous outage on February 11, 1986. Due to transmitter cover threads being galled the transmitter was replaced with a bench calibrated spare. This condition, therefore, existed during periods of operation in March, April, and May 1986 when the AFP was required to be operabl . .. - - . .

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Administrative Procedure 5.04, " Control of Installed Plant Instrumentation (IPI)" Section 7.4, requires that upon return to service of any IPI that is not associated with an indicator be verified using appropriate test equipment for verification. Contrary to these requirements Loop verification was not performed on the replaced transmitter. This constitutes a violation of Technical Specification 6.8.1 as set forth in the Notice of Violation, Item b (255/86023-01(DRP)).

As documented in Event Report E-PAL-86-075, corrective action has been taken to assure that proper instrument loop verification is performed after calibration. The verification is documented on the Work Order and the appropriate supervisor verifies proper completion. Loop verifications were performed satisfactorily on the remaining AFP suction pressure transmitters. These corrective actions appear adequate to resolve this violation and the inspectors have no further questions regarding this matte Starting on August 4,1986, a series of discoveries were made related to the operability of containment air cooler VHX-3. On that date the licensee found that the service water outlet valve to the containment air cooler was not operating correctly. At that time it was thought that the valve failed closed during actuation of a Safety Injection Signal (SIS). It was later determined that the valve actuator was adjusted such that the valve did not fully close nor fully open by about 22% of full travel. This condition was also found to exist on containment cooler VHX-2 which is a part of the same engineered safeguards train. This reduction in flow is considered sufficient to render the two containment air coolers inoperable. Since review of the maintenance history information regarding the cooler outlet valves and their actuators has not disclosed an occasion where any adjustment would have been performed, it is presumed that the condition has existed since original plant installatio Previous operation of the plant with these components inoperable is prohibited by Technical Specification 3. The above discovery was made as a result of a system flow test performed in March 1986 at the end of the refueling outage. This test noted flow variance between the coolers which required further investigation. This was one of the actions committed by the licensee to be completed under the Material Condition Task Force revie The licensee's evaluation of the safety significance is summarized as follows: The limiting accident for containment neat removal (loss of coolant) requires a heat removal rate of 2.29 E8 BTU /hr from the three containment air coolers associated with Emergency Diesel Generator 1-2. Using the flows measured in the March 1986 flow test which are indicative of the discovered condition, a total heat removal rate of 1.62 E8 BTU /hr has been calculated for Containment Air Coolers VHX-1, VHX-2, and VHX-3. The 6.70 E7 BTU /hr discrepancy is more than compensated for by the additional 1.20 E8 BTU /hr heat removal capability from Containment Spray Pump P-54A,

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which is also required to be operable per Palisades Technical Specification 3.4.1(a), but for which no credit is taken in the FSAR analysis. The licensee concluded that in spite of the reduced service water flow through the containment air coolers, adequate accident heat removal capability existe This e' vent will be reviewed in additional detail by the NRC and may be the subject of escalated enforcement action. The licensee is still conducting system testing after valve adjustment and is also evaluating system flow requirements. Evaluation of this event in conjunction with the missing access plate on VHX-3 which reduced air flow (Paragraph 3.j) has not been completed. Resolution of this issue will be tracked as an Unresolved Item (255/86023-02(DRP)).

e. An evaluation of the Component Cooling Water (CCW) system was conducted as part of the licensee's investigation into a safeguards pumps cooling water low flow alarm which is a Material Condition Task Force item. On August 7, 1986, the licensee determined that the Engineered Safeguards pumps were not getting the 140 gpm CCW cooling flow listed in FSAR Table 9-7. The assumption is 5 gpm to each containment spray (CS) pump, 10 gpm to each low pressure safety injection (LPSI) pump and 35 gpm to each high pressure safety injection (HPSI) pump. The 140 gpm total would also include the original third HPSI pump which has now been converted to a third auxiliary feedwater pump. The investigation to date has determined that of the components mentioned above only the HPSI pumps do not get adequate flow. Joint licensee and vendor evaluation concluded that the loss of or reduced CCW flow to these pumps would not affect operability until approximately 3800 hours0.044 days <br />1.056 hours <br />0.00628 weeks <br />0.00145 months <br /> of operation since CCW cooling was provided to extend bearing and seal lif During preliminary test runs to evaluate (CCW) flow on August 18 and 24, 1986, the licensee discovered that the CCW system may be incapable of supplying the flow listed in FSAR Table 9-7 to the Shutdown Cooling Heat Exchangers (SDCHX) during containment sump recirculation cooling. The outlet valves on the CCW heat exchangers (CCWHX) have been throttled since approximately 1971 to limit vibration induced damage to the CCWHX. In the CCW throttled condition the CCW flow to the SDCHXs is 5800 gpm while in the unthrottled condition flow is 7950 gpm. FSAR Table 9-7 states 8000 gpm. After additional evaluation, the licensee made the 10 CFR 50.72 notification on August ?5, 1986, at 4:21 In September 1986, the licensee discovered documentation that the CCWHXs were sized at half of the capacity specified on the purchase specification. The licensee wanted heat exchangers sized for 5700 gpm each. The actual heat exchangers supplied are for 5700 gpm total. The documentation supplied with the heat exchangers specified 5700 gpm per unit. The vendor defined a unit to be two shells. The licensee apparently missed this definition and assumed a unit to be one heat exchanger, therefo:e, one shell. Preoperational

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testing did obtain 5700 gpm through each heat exchanger but experienced tube damage from excessive flows. The licensee then limited differential pressure or flow to prevent tube damage in the futur The licensee plans to resolve the CCW flow problems through engineer-ing analysis and testing to determine what flow is adequate to meet cooling needs. The issue of incapability of meeting the FSAR required CCW flow to the SDCHXs will remain an Unresolved Item (255/86023-03(DRP)') ~

pending further NRC revie f. On August 8,1986, the licensee discovered during an evaluation for the addition of future loads to the nonvital 1E bus that the feeder cables to the vital busses 1C and ID as well as the non-vital 1E bus may be overloaded as documented in Deviation Report D-PAL-86-19 The limiting routing method is for conduit exposed to sunlight which is used for cabling from the station power and startup power transformers to the IC, 10, and 1E buses. The conduit in sunlight capacity rating is 859 amps while the calculated Loss of Coolant Accident (LOCA) maximum loading is 919 amps. The 1E cable from the 1-1 station power transformer has experienced loadings in excess of 859 amps on approximately 15% of the 625 days the 1E bus was being fed from the station power transformer during the past two and one-half years. It has also experienced several days of loading above 950 amps. Based upon the actual loadings experienced on the IE cable which is of the same type as the 1D cable and the conservatism involved in sizing the cable, the licensee believes the cable would not fail during a LOCA. This situation does nat affect the cable from the diesel generator to the 1D bus. The licensee plans on installing solar shields to improve the ampacity of the feeder cable to the ID bus and improving the ampacity of the feeder cable to the IE bus by installing covered cable tray As part of the corrective actions, the licensee plans on doing a calculation to determine the short term rating of the cabling and establishing better controls for adding loads to the buses to prevent overloading of any portion of the system. Open item 255/86023-04(DRP) will be used to track completion of the licensee's corrective action g. On August 14, 1986, after completing maintenance on a coolant charging pump (P-55B), the licensee performed a test run of the pump. The Shift Supervisor considered it fully functional without performing the Technical Specification (TS) surveillance tes l TS 3.2.1 requires "when fuel is in the reactor, there shall be at least one flow path to the core for boric acid injection." Being a .

new shift supervisor he compared that TS with 3.2.2 which requires !

that "at least two charging pumps shall be operable" for critical l operation. He concluded that since the word " operable" was omitted I from TS 3.2.1, he did not have to declare the pump operable before removing the other pump (P-55C), which had been used for the boric i acid flowpath, from service. Review by a subsequent shift identified j l

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the problem and P-55B was declared operable at 4:05 a.m. on August 15, 1986. A 10 CFR 50.72 report was made at 6:42 a.m. on August 15, 198 To prevent similar interpretations, the licensee is planning to make procedural change h. On August 15, 1986, at 4:48 p.m. the ID vital bus was deenergized and both diesel generators (DGs) started due to an inadequately controlled maintenance activity for relocating auxiliary fuses in a station power breaker. The fuses were being moved to reduce interference when the breaker was pulled from the cubicle. This work was not a result of the new action plan. While performing the Q-listed Work Order No. 24606289, a technician opened three links to check out the wiring modification and subsequently did not close them. The work order did not indicate that the links were to be opened or left open. When the potential transformer fuses were reinstalled the second level undervoltage relays sensed no voltag When the knife switch to the undervoltage relay was cut in, bus ID undervoltage was received. The startup power breaker tripped open, loads were shed and both diesel generators started as designe The 10 CFR 50.72 notification was made at 8:07 During event review the licensee discovered that the 1-2 diesel generator breaker did not automatically close onto the ID bu A wiring error was discovered involving the bus undervoltage scheme relays that prevented the DG breaker from closing. What was found was that the load shed relays were powered from another scheme for station power that had been deenergized for main transformer replacement. This prevented the load shed check interlock from functioning to allow the DG breaker to close. This was also found to be the case on the vital IC bus schem Maintenance is planned to correct the wiring error on both buses, but the wiring error does not affect system operability during normal plant electrical alignment. This event is under review by the NRC and may be the subject of escalated enforcemen On August 23, 1986, at 1:35 p.m. a similar automatic start of the diesel generators occurred with the loss of the vital IC bus. The licensee was preparing for the same modification as noted above on the IC bus. The tagging and Switching Order No. 86-1414 that was prepared did not specify that both undervoltage relays must be cut out; only one was. When the potential transformer drawer for the 1C bus was pulled, undervoltage was sensed by the relay that was not cut out and a load shed sequence was initiate The feed from the startup transformer tripped open, deenergizing the IC bus and both DGs started. Due to the above noted existing wiring error, the 1-1 DG breaker did not automatically close onto the IC bus but was manually closed. The 10 CFR 50.72 notification was made at 3:28 On August 27, 1986, at 12:54 p.m., a third inadvertent diesel generators automatic start actuation occurre Instrument and Control (I&C) technicians opened a link to block automatic diesel

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l generator (DG) starts prior to commencing turbine thrust bearing l trip device replacement to avoid inadvertent DG starts. The maintenance was being done to correct a Material Condition Task Force item of clogged orifices in the trip system. The link was reclosed seconds later when the I&C technician erroneously thought the state of some nearby relays was incorrect and attempted to restore the circuit to its initial condition. The links were reclosed without prior notification to the control room operators who could have blocked the DGs starting from the control room as part of the restoration proces The 10 CFR 50.72 notification was made at 2:02 p.m. The opening and closing of the link was not specified on the Q-listed Work Order TGS 2460663 I&C in conjunction with the Shift Supervisor decided to reposition the link and document same on Work Order.

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In the three events discussed above inadvertent diesel generator starts and in two cases loss of vital buses have occurred in part due to inadequately controlled maintenance and modification activities. The August 14, 1986, event was due to lack of specific instructions in the work order to ensure proper restoration of the system following the modification activit The August 23, 1986, event was due to inadequate preparation of the tagging and switching order since it'did not isolate all affected components. This tagging was done in accordance with instructions provided by Electrical Engineering to perform the work. Several knowledgeable reviewers had missed the error. The August 27, 1986, event was due to a personnel error in judgment but the activity was inadequately controlled in that the opening and closing of the link was not addressed in the work order.

l Another ramification of these events is the lack of independent verification that Q or Safety related activities receiv ANSI N18.7-1976 Section 5.2.6 " Equipment Control" requires that

". . . procedures shall require independent verifications, where appropriate to assure that necessary measures, such as tagging equipment, have been implemented correctly. Temporary modifica-tions, such as temporary bypass lines, electrical jumpers, lifted electrical leads, and temporary trip point settings shall be controlled by approved procedures which shall include a requirement for independent verification. . ." Controls of the lifting of leads which are part of maintenance orders do not receive the same level of control as those controlled by formal procedures. .The lack of specific procedures and reviews, and the lack of independent verification requirements are considered contributors to the above event Administrati.e Procedure 5.01, " Processing Work Requests / Work Orders," Attachment 4A, Block 46 through 50, which describes the job plan, requires that maintenance which can affect the performance of Q-listed equipment, structures, or systems, must be properly preplanned and performed in accordance with written procedures,

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instructions, or drawings appropriate to the circumstances. The three examples of inadequate work instructions is considered a violation of Administrative Procedure 5.01 as set forth in the Notice of Violation, Item a (255/85023-01(DRP)).

i. On August 18, 1986, at 5:08 p.m. a right channel safety injection signal (SIS) occurred during a modification tes Equipment available for service responded as expected. Three component cooling water (CCW) valves had been modified and the post modification test procedure for the CCW containment isolation valves logic manually made up the relay which caused the SI The procedure stated that a right channel SIS would occur but the shift operations staff convinced themselves that an SIS would not occur due to SIS being defeated in the current plant condition.

i Only the low pressurizer pressure SIS was defeated and the test made up the containment high pressure SIS. The licensee made a 10 CFR 50.72 notification at 7:34 p.m. for an unplanned ESF actuatio j. During investigation as to the cause of a Containment Air Cooler fan motor (V-1A) failure in July 1986, it was discovered than an access

, plate to the ducting between the fan (V-3A) discharge and the cooler

! (VHX-3) was off. After an engineering analysis was completed in

! August, a 10 CFR 50.72 report was made on August 25, 1986, reporting l that the fan would not have met its design air flow requiremen The licensee has not reached a final conclusion as to how the access cover was removed, but currently feels that the cover retaining bolts may have vibrated off. Many structural bolts were also found to be loose or missing, contributing to this theor The V-3A fan motor bearings were found failed, and a winding was found to be open circuited. The licensee believed that the open winding was the source of vibration which loosened the bolts and ca'used the motor bearing problem. The bypassing of air flow rendered the cooler inoperable for an indeterminate period of time, possibly for several years of plant operation. This condition is in violation of Technical Specification 3.4.1. The licensee is continuing to evaluate the combined safety significance of the VHX-3 containment air cooler service water flow and air flow being less than design '

requirements. This issue is a candidate for escalated enforcement and is also being tracked under Unresolved Item (255/86023-02(DRP)).

A Licensee Event Report is due on this aspect of the cooler problems i on September 25, 198 ;

The cause for the V-1A fan motor failure ~was a failed damper whose welds had broken allowing it to remain closed. All similar damper welds were inspected and 67 were found in need of repai These repairs were made by the licensee. No surveillance of the coolers or fans is required by the Technical Specifications and apparently

, none was being done. Licensee corrective actions are being finalized to prevent recurrence and include routine outage vibration monitoring and air flow surveillance testin _____-___ _ - _ _ _ _ _ _ _ - _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - _ _ .

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pump performance during surveillance testing, the licensee decided to rebuild the P-678 pump. This action was carried out under the licensee's Material Condition Task Force commitment. After rebuilding and testing it was found that at the design discharge pressure (350 Foot TDH) the pump delivered 2700 gpm instead of 3000 gpm as required by the FSAR and Safety Analyse The NRC was notified of this discovery on August 25, 1986, by 10 CFR 50.72 repor Licensee communication with the pump vendor determined that prior to initial operation in 1970, the "B" pump was modified to improve its performanc Later the "A" pump was similarly modified. In 1983, the "B" pump rotor seized and was replaced with a spare. The spare rotor had not been similarly modified and at the time there were no records on site indicating that it should have been. Post maintenance testing did not ensure conformance to system design requirement Noting that the P-67B pump performance did not compare with previous ASME Section XI test results, the responsible engineer established a new baseline and informed the system engineer of the discrepanc No action was taken to investigate at this tim Thr. licensee rebuilt the pump with a properly modified impeller and retested the pump with satisfactory results. Plant operation with the pump inoperable during periods of critical reactor operation between July of 1984 and May of 1986 was in violation of Technical Specification 3.3.l This issue remains an Unresolved Item (255/86023-05(DRP)) pending NRC review for escalated enforcement and further review and analysis by the license . During a thunderstorm at 10:33 a.m. on August 26, 1986, two switchyard breakers opened, apparently due to lightning strikes. One of the two automatically reclosed and no power was lost to the plant which was in cold shutdown. Two minutes later vital ID bus feeder breaker from startup power tripped causing an undervoltage condition on the ID bus and automatic start of the diesel generators. Due to system alignments for main transformer work, the 1-2 diesel was manually placed on the ID bus restoring the associated service water pump and shutdown cooling pump. The 1-1 diesel generator was shutdown since it was not needed. At 11:15 a.m. nonvital bus 1E tripped when the potential transformer drawer containing the IE bus undervoltage relays was inadvertently pulled during troubleshooting activities for the loss of the ID bus. Apparently a lightning strike caused the feed from the startup transformer to the ID bus to trip open and blew the potential transformer fuses for both the ID and IE busse Following replacement of those fuses, the busses were re-energized from startup power by 1:30 During simulator training, operators identified a potential for the loss of both trains of Emergency Core Cooling Systems (ECCS). With a DC bus de-energized including both of its preferred AC buses

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(which provide control power) an SIS would occur, but only the

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opposite train's pumps and valves would actuate with control power lost to the other train. Also, a sump recirculation switchover would take place causing the operable ECCS pumps to line up to the empty sump. It was postulated that a dead short to ground on the DC bus could activate the sequence. This was reported under i 10 CFR 50.72 at 4:30 p.m. on August 26, 1986. A modification is planned to resolve this condition which changes the two out of four

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logic to one out of two taken twice. This separates the power divisions so that a DC bus failure would not cause a recirculation actuation signal nor prevent a valid signal with this failur It was found that this condition also existed at other plants with automatic sump switchover. The modification will be done under a 10 CFR 50.59 evaluation performed on September 5, 1986. The licensee may need to change the basis for Technical Specification 3.1.7 and other Technical Specifications where the logic is referred to as being two out of four, as well as make other procedural changes and conduct operator training. The completion of the mcdification controlled under Facility Change FC-707 will be tracked as an Open Item (255/86023-06(DRP)). As required by Technical Specification 3.6.5a Containment Air Room Purge Valves CV-1813 and CV-1814 are to be electrically locked closed during hot shutdown, hot standby, and power operatio During engineering review of a different issue and subsequent testing by operations during cold shutdown on August 30, 1986, it was determined that starting the purge fan V-46 opens the isolation valves even when electrically locked closed. Containment high radiation or high pressure signals both override the fan start opening of these valves. This fan interlock feature was unknown by operators whose procedures have required opening the air valves prior to starting the fan. The licensee is not aware of any

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situations where the fan was started to defeat the locked closed

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q requirement. Technical Specification 3.6.5a allows one hour to reclose the open valve and six hours to be in hot shutdown. The valves are verified to be in the correct position on each shift turnover. The licensee plans to modify the circuitry to accurately reflect the Technical Specification requirements. Completion of this modification will be tracked as an Open Item (255/86023-07(DRP)).

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One violation with several examples was identifie . Maintenance The inspector reviewed and/or observed the following selected work activities and verified whether appropriate procedures were in effect controlling removal from and return to service, hold points, verification testing, fire prevention / protection, and cleanliness:

Design Basis Accident Sequencer No. 34-1 (Maintenance Procedure ESS-I-13)

Rebuild of the P-55A Fluid Drive (Work Order CVC-24606028, SC 86-200)

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Rebuild of oil system on P-55A (Work Order CVC-24605776, SC 86-175)

Modification to Escape Lock Cam (SC-86-216)

No violations or deviations were identifie . Surveillance The inspectors reviewed surveillance activities to ascertain compliance with scheduling requirements and to verify compliance with requirements relating to procedures, removal from and return to service, personnel qualifications, and documentation. The following test activities were inspected: MI-39 Auxiliary Feedwater Actuation Logic Test HP Process Monitor Operational Check-Quarterly D/WO-1 Daily Control Room Surveillance No violations or deviations were identifie . Engineered Safety Features Walkdown The inspector performed a walkdown of the Boric Acid Flowpath and verified:

That each accessible valve in the flowpath was in its required position and operable, that power was aligned for components that activate on an initiation signal, that essential instrumentation was operable, and that no conditions existed which would adversely affect system operatio Several valves that should have been locked in position were found in the correct position but the locks were not closed. The Shift Supervisor was informed and had them locked. This is an example of a poor practice of not strictly controlling valves in a required flowpath during a shutdown condition and was discussed during the Management Interview with the license No violations or deviations were identifie l 7. Licensee Event Reports l

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Through direct observations, discussions with licensee personnel, and review of records, the inspectors examined the following reportable i events to determine whether: reportability requirements were met; immediate corrective action was accomplished as appropriate; and corrective action to prevent recurrence has been acco:aplished per Technical Specificatio I (Closed) LER 255/86011 - Revision 1: During the month of February, while in cold shutdown, a high startup rate reactor trip occurred due to instrument noise generated by a welding activity near the location of a nuclear instrument. Several days later while in hot shutdown, a high startup rate trip occurred due to instrument noise generated by a

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chattering relay in a radiation monitor. Noise generated from welding operations is expected but the licensee currently appears to have gained more control over welding activities since these February events.

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Previously the licensee planned on replacing the chattering relay but subsequently reinstalled it for troubleshooting and has observed no similar problems. The suspected cause was an inadequate connection between the relay and the socket. The connections were cleaned and the problem apparently solve (0 pen) LER 255/86022 - Revision 1: Evaluation of two-inch diameter piping in the High Pressure Safety Injection (HPSI) System using the existing FSAR Seismic Response Criteria concluded that the piping would be in a calculated overstressed condition during a postulated seismic

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event. These eight lines comprise all of Train 1 and Train 2 HPSI flowpaths to the Primary Coolant System; therefore, during a postulated seismic event HPSI could be rendered inoperable. This condition has existed since original plant constructio The evaluation was prompted wnen a motor operator on one of the HPSI valves was to be replaced with a heavier operator. The HPSI lines were not evaluated as part of the IE Bulletin 79-14 or SEP Topic III-6 reviews which included small bore piping system analysis on a sampling basis. The identified overstress condition has been analytically resolved by the licensee with no additional supports being added to the HPSI system by utilization of 1 ASME B&PV Code,Section III, Code Case N-411. This code case permits a significant increase in seismic response spectra damping values for piping analyses which would predict a lower subsystem response for a given set of plant spectral curve For Palisades, the result is a lesser seismic input and smaller resultant seismic stresses which yield acceptable results for the HPSI lines. A submittal was made by the licensee to Nuclear Reactor Regulation (NRR) on July 28, 1986, notifying NRR of the use of Code Case N-411 and NUREG/CR-1833 Seismic Response Spectra for Palisade The FSAR will be updated during the next updat l This LER will remain open pending NRR's review of the submitta (Closed) LER 255/86023: Containment escape lock design deficiency was l discovered and discussed in Faragraph 3.b of this repor l (Closed) LER 255/86-024: The Containment Air Coolers VHX-2 and VHX-3 were found to have their service water outlet valves throttled. This event is discussed in this report, Paragraph No violations or deviations were identifie : Followup on IE Bulletins (Closed) Bulletin 84-02 " Failure of General Electric Type HFA Relays in use in Class IE Safety Systems" directed inspections and replacement of

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approximately 177 relays at Palisades. Prior to startup from the then current refueling outage, the licensee replaced all normally energized relays, thus averting the need to inspect or provide a justification for continued operation. A review by the licensee determined that these were the only relays used in safety-related applications. The licensee's

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July 13, 1984, response to the Bulletin was three days late, but since corrective actions to replace the energized relays weic already complete, this error is viewed as insignifican The licensee subsequently updated their response on October 11, 1985, to amend their schedule to complete safety-related relays to correspond with the end of the 1985-1986 refueling outag (Closed) Bulletin 86-01 " Minimum Flow L.ogic Problems that Could Disable RHR Pumps" was issued May 23, 1986, to all G.E. boiling water reactor This bulletin, concerning G.E. design deficiencies, is not applicable to Palisade (Closed) Bulletin 86-02 titled " Static "0" Ring Differential Pressure Switches" was issued on July 18, 1986, and required initial action within seven days. Consumers Power responded on July 24 that none of the subject differential pressure switches are ". . . installed (or planned)

as electrical equipment important to safety, as defined in 10 CFR 50.49 (b)." Therefore, no further action was required. The inspector reviewed the licensee's method of determining that they had none of the switche . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 3.d, e, j and . Open Items Open Items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or bot Open items disclosed during the inspection are discussed in Paragraphs 3.f, m and . Management Interview A management interview (attended as indicated in Paragraph 1) was conducted on September 8, 1986, following the inspection. The scope and findings of the inspection were discussed. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.

, The licensee did not identify any such documents / processes as proprietar ,

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