IR 05000255/1996017

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Insp Rept 50-255/96-17 on 961124-970110.Violations Noted. Major Areas Inspected:Licensee Operations,Maint,Engineering & Plant Support
ML18066A891
Person / Time
Site: Palisades Entergy icon.png
Issue date: 02/20/1997
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
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ML18066A889 List:
References
50-255-96-17, NUDOCS 9703030121
Download: ML18066A891 (21)
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Text

U.S. NUCLEAR REGULATORY COMMISSION Docket No.:

License No.:

Report No.:

Licensee:

Facility:

Location:*

Dates:*

Inspectors:

..

Approved by:

REGION Ill

  • consumer:s Power Company 212 West Michigan Avenue Jackson~ Ml 49201

.

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Palisades Nuclear Generating Plant 27780 Blue. Star Memorial Highway

  • Covert, Ml 49043-9530 November 24, 1996 through January 10, 1997 M. Parker, Senior Resident Inspector
  • P. Prescott, Resident Inspector
  • Bruce L. Burgess, Chief Reactor Projects Branch 6

. EXECUTIVE SUMMARY Palisades Nuclear Generating Plant NRC Inspection Report 50-255/96017 This inspection reviewed aspects of licensee operations, maintenance, engineering and plant support. The report covers a 7-week period of resident inspectio *

Ooerations

The inspectors concluded that appropriate control of primary coola11t sys~em (PC_S)

temperature. was.not provided to ensure that PCS temperature was maintained-. :*

above 525°F. * $everalfactors contributed to this, most notably, the low:decay.

heat1 -complicated by the urgency of a quick shutdown due to a steam lea!:< on the main steam isolation valves (MSIV's) and operator over r*eliance on the simulat_or~ * *

.There was a lack of sensitivity to the need for additional c;ontrol.and oversight of*

PCS temperature. * A violation of technical specifications was identified* for the failure to maintain PCS temperature abov_e 525°f while the reactor.was crit_ical (Section 01.2).

  • Although two core exit thermocouples were cross CQnnected during the outage by contractor. maintenance personnel, the licensee identified this c_onqition in. _a ti.mely and appropriate manner through_ *routine surveillance activities. The *1icen.see *

subsequentl_y'-~ook action to correct this condition.during aJorced. outage.:* One.

unresolved :item was* identified concerning. the cross connected.thermocouples-:

  • (Section_ Of.3).

Th~ inspectors concluded that the.licensee did not provide adequatt;t oversight of *

the core: operating parameters, resulting in exceeding specified admil"!istrative limits :

for radial peaking factors and potentia_lly technical specification.1.imits.. This wa...

considered an* unres.olved item pending-further review by both the licensee.and the **

NRC into the licensee's revised calculations (Section OL4).. *

Weaknesses were n.oted in the licensee's procedures for verifying c*ontainment *.*.

cleanliness following. work performed after th_e containment. closeout' walk~thro~gh':

was completed (S~ction 01.5).

Maintenance.. :

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The inspectors viewed the overall licensee performance during the refueling outage*

.as good. Many longstanding equipment problems were addressed (Section M1 ~2).

.. 2

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There were several equipment problems involving rework which._necessitated two forced outages. Problems were encountered with weld repairs on both main steam.

isolation valves; cracking of two main generator isophase connectors, which was

  • similar to a December *1995 event; leakage from the "D" primary coolant pump seal
  • stage; and necessary re-adjustments to the mechanical overspeed trip linkage and the governors for both main feed pumps (Section M1.2). *

The licensee's attempt to identify potential equipment deficiencies that may not.

_meet th~ guidance of GL 91-18, was found to be thorough. However, the inspectors identified examples of weaknesses in followup on corrective action of the identified deficiencies to assure that problems were resolved either by repair of the affected equipment or that an adequate engineering safety*analysis was performed (Section fv'l1.3).

Engineering

  • *

. A major emergent issue that arose during the refueling outage was the

  • environmental qualification (EQ) of Rome cable splices in the containment. The splices had been improperly performed.. Planning and execution of the containmen~

Rome-splice issue was good. However, weaknesses were.noted.in determining the initial scope of the problem (Section E1.1 ).

  • *Plant Support

The inspectors observed during daily plant walkdowns that* radiological" worker practices during the refueling and maintenance outages were adequate (Section R1.1 ).

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I

REPORT DETAILS Summary of Plant Status

  • -A 56-day refueling outage was compieted on December 27,_ 1996. Due to cracking of two main generator isophase bus connectors, the unit entered a forced outage on January 5, 1997. The connectors were repaired, and the unit was returned to service on January 6, 1997. However, both. main steam isolation valves developed leaks at the plugged leakoff line on the valve actuators, and power reduction was commenced on *January 6, 199 The unit was placed in cold shutdown on January 8, 1997. During the second forced outage~ the "D" primary coolant pump seal was replaced due to pressure oscillations noted in one *of three rotating seal stages. The unit _was returned to service January 14, 1997, reaching full power on January 17, 199 * I. Ocerations

Conduct of Operations O 1. 1 General Comments (71707)

Using lnspectio_n Procedure 7*1707, the inspectors conducted frequent revie~s. of ongoing plant operations. The conduct of operations was good; specific events arid noteworthy observations are detaile~ belo.2. Primary Coolant System <PCS> Temperature. Less Than 525°F a... Inspection Scope. <71707)

The i!lsp~ctors reviewed the events leading up to the deerease in primary coolant

  • . temperature to._less than *525°F on January 6, 1997, following a plant Startup anct'a
* * : : * : subsequent plant shutdow'1**
  • *b.: *: Observations and Findings 1.:- * On January 6, 1997, the licensee was preparing*to place the.turbine on line during.*.

. r'ecovery *from a forced outage the previous day. *PCS temperature was increased.. *

to 536°F~ approximately.4°F above no-load average primary coolant system temperature (T eve>, resulting in the bypass valve opening approximately 75 percent..

After piacing the turbine on line, turbine load was increased to 6 percent. This *

' action increased.steam demand, resulting.in a decrease in.T.v.* PCS temperature decreased-below 525°F for approximately 47 seconds during the evolution, dropping *to 524°F. Technical Specifications 3.1.3 a) required* that the PCS temperature. be maintained above 525°F while the reactor*is critica.

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.

In reviewing the circumstances leading to the PCS temperature drop, the inspectors noted that the operators had received just-in-time. training on the simulator for..

startup activities. During this training~ the operators evaluated how much control

. rod rflotion would be needed when placing the turbine on-line. The shift concluded

..

that 3-4" of rod motion was required; however, approximately 18" of rod motion was actually required while plaCing the turbine on-lin The licensee evaluated several potential contributors to the decrease in PCS.

temperature:

Operator over reliance on simulator respons *

Discrepancy between required rod motion between simulator and actual plant. *

The failure to maintain PCS temperature above 525°F was considered a violation (50-255/96017-01 (DRP)) of Technical Specifications 3. i.3 a). On January 6, 1997, while the licensee was preparing to shutdown the reactor, PCS temperature again decreased below 525°F. The operators were taking the. turbine off-line, and the bypass valves picked up additional steam load. One of the bypass valves was noted by plant operators to be reacting erratically to control system demand. Recognizing that PCS tempe.rature was decreasing, the operators secured the* main feed pump and subsequently closed the main steam isolation valves to reduce steam demand. PCS temperature dropped below 525°F. for appr.qximately five minutes, dropping to 522°F before recovering to 525°F. *

The licensee's initial evaluation noted that a main contributor to exceeding the Technical Specification limit was the relatively low decay heat generated by the reactor due to the limited power history of the reactor plant (on line less than eleven days), Additionally, the impact of the turbine bypass valve was evaluated.*

The valve* was *tound with its packing f_ollower backed-off and one layer of *packing

  • blown out. This condition could have caused the erratic or sluggish operation* of
  • the bypass 'valve discussed earlie *

Further review by the.licensee determined that the reactor was sub-critical during the time that the PCS temperature dropped below 525°F: This determination,.

based on a reactivity balance performed by reactor engineering sut>sequent to the power excursion, was performed as part of the licensee's condition* report investigation for this inciden Conclusions The inspectors concluded that appropriate control of PCS temperature was not

  • provided to ensure that PCS temperature was maintained above 525°F. Several factors contributed to this event; most notably, the low decay heat, complicated by the urgency of a quick shutdown due to a steam leak on the MSIV's; and operator over reliance on the simulator. A violation *of TS was identified for the failure to maintain PCS temperature above 525°F while the reactor was critica ***

01.3 Core Exit Thermocouoles Inspection Scope (7170.7 and 617261 On December 29, 1996, following a plant startup from the refueling outage, the licensee declared core exit thermocouple (CET) #10 inoperable. The inspectors observed the licensee's activities to return CET #10 to servic Observations and Findings On December 29, 1996, during the performance of a Palisades lncore Detector

  1. 10 was not meeting its surveillance acceptance criteria. The CET was responding to flux changes; however, it was reading approximately 30°F lower than expected for its core location and associated power leve Further review by reactor engineering identified that CET #13 also did not appear to
  • be trending as expected for its associated core location, although it met acceptance criteria. Subsequent evaluation of the readings of the two detector strings indicated that the connectors for CET #iO and CET #13 were swapped during installation* of the reactor vessel head. CET #1 O was considered a qualified detector, but CET #13 was a non-qualified detector, so the licensee took appropriate action and declared CET #10 inoperable per Technical Specification (TS).

3.17. This resulted in* the licensee entering a 48-hour TS Limiting Condition for Operation (LCO) *due to having two required CET channels inoperable in the.same core quadrant. During reactor vessel head assembly, CET #19, a qualified detector, was found inoperable due to a cracked weld on the detector fitting. Both CET #10 and # 1 9 were located in the same core quadran The licensee performed an operability assessment and concluded that although CET

  1. 10 and CET # 1 3 were cross-connected* at the reactor vessel head connection, the detector string for CET #13 was fully qualified and met environmental qualification
  • requirements. The licensee had procured all detectors as safety-related; therefore, all incore.detectors met safety-related quality standards. Additionally, the reactor vessel head connection was considered environmentally qualified because the connectors were replaced during the outage with safety-related connectors. On December 30, 1996, the licensee declared CET #10 operable and subsequently exited the LCO. As of. the end of the inspection period, the inspectors were still.

evaluating the issue of how the nonsafety-related detector string could become cross-connected with a safety-related incore string. This issue will remain unresolved pending completion of additional inspectio Conclusions Although the connectors were cross connected during the outage by contractor

  • maintena_nce personnel, the licensee identified this condition in a timely and appropriate manner through routine surveillance activities. The licensee su_bsequently took action to correct this condition during the forced outage that

',*...

  • commenced on January 6, 1997; *however, an Unresolved Item (50-255/96017-02 (DRP)) was identified concerning the cross-connected thermocouples pending completion of additional inspectio.4 Radial Peaking Factors Inspection Scooe <71707 and 617261 On December 31, 1996, during a routine review of the Palisades lncore Detector Algorithm (PiDAL), a reactor engineer identified.that the maximum total pin peaking factor value of 1.957 exceeded the Technical Specification (TS) limit of 1.954. The
  • inspectors reviewed the events leading up to the excessive *Radial Peaking Factors following a plant startup from the refueling outag * Observations and Findings On December 30, 1996, the licensee performed* a power escala.tion to full power following a startup from the refueling outage on December 2~, * 1_996. At'

approximately 93 percent power~ a Pl DAL run* cpnf_irmed that the core operating*

parameters were within limits. The reactor engineer evaluated th.a data and determined that it was acceptable to proceed to full power with no furthe *.

.. *.

. restrictions. Based *on the acceptable PIDAL run* at 93 percent power> no furtft'3r -- *

evaluation was deemed necessary. Full power was reached on* December 3 ~,

. 1996. Appr~ximately six hours later, reactor engineering reviewed the PIDAL and.:

determined that.the radial peaking factors for fuel assembly 0:-06, in core location H-20, may have exceeded TS limit *

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TS 3.23.2, required that with reactor* power greater than 50 percent of rated power*

and any radial peaki.ng factor exceeding its limit, reactor power shall be redu.ced *

within 6. hours. The control room was informed* of this condition and* re~uced * -*.. :.

  • * reactor power to within the TS limits. Reactor power was reduced to 9? percent. *

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The inspectors* evaluated this event for causal factors and identified the f ollowirig. * *

contributors. For a *period of approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, a reactor engineer :was not :

pre~ent to evaluate radial peaking factors.. * Although a reactor engineer was on_site to evaluate the. 93.percen.t PIDAL run, the Inspectors concluded that the decision to proceed to full*.power based on PIDAL data taken at 93 percent was not a conservative one: The decision was non-conservative because radial peaking

  • factors were within 3 percent of the Technical Specification limits and* were higher than the 'predicted values based on the fuel vendor's evaluation. The licensee planned ~a.reevaluate the requirements for reactor engineering site coverage for major power evolution The licensee c;orisulted with the fuel vendor and preli~inary cal.culations determined that the radial peaking factors might not have exceeded TS limit.s. * The radial peaking factors werEtbased on achieving full power at 150 megawatt days per*

metric toA burnup (5 effective full power days). The plant actually achieved full power in a much shorter time period due to a short chemistry hold* and no required

. t

. reca.libration of nuclear instruments. The fuel vendor and the licensee were reevaluating the radial peaking factors based on the* actual fuel burnup*.". *

Conclusions the inspectors concluded that the licensee provided limited reactor engineering *

oversight of core operating parameters during initial plant startup from an c;>utage.

. Limited coverage by reactor engineering contributed* _to exceeding administrative limits for radial s)eaking* factors. This was *consid~red an Unresolved Item* (50-.

255/96017-03 (DRP)) pending further review of the. licensee's revised caiculati9ns by-both the licensee and the N~C. *

01.5 Control of Containment After Closeout W~lk-ThrOugh* lnsoection Scope (71707>

The. licensee ide_ntified additional work which required containrne"'!t*e~fry*atter the closeout walk-:through was completed.- The inspectors reviewed *th~ licensee's pr9Cedures for containment entry control above cold shutdown.*

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. *Observations and Findings The licensee had determined that the cable on a* seniice water valve to containme~t a*ir cooler needed to be replaced.-: This job required scaff~lding*;

however, Operations was not notified that scaffolding was erected in containm~nt. *.

to perform the task. _Procedure GOP-2, "Plant Heatup <Cold Shutdown ~o Hot * *

Shutdown)," containe.d a requir~ment to per.form Checkli~t 1.4, "Containment.

Closeout Walk-Through," prior to. exceeding* 210°F (cold shutdown). The. proc~*dure.

did not have a statement to control containment entry after the walk-thrqugh was. : : :

complete~ Administrative procedure 4.02, "Control of Equipment,". contained a.

step requiring the card* reader to the personnel airlock to be de-energized when the *

  • primary systel'TI was above cold shutdown. This pr~cedure *also required *~he...

contaim:nent entry to be logged in the shift supervisor's logbook if a containment entry was required above cold shutdown. Although plant *proc~dures were not violated,* neither procedure contained measures to control work performed in containment after the checklist was completed but before the plant was taken *

above cold.shutdow *

T_he inspectors discussed the above procedural weaknesses with a*ppropri~te

. personnel. The licensee stated that they would review this issue and evaluate the *

need for procedural changes to address th~ inspectors' concern* for contrql of containment entries to perform work after the containment cleanliness checklist, CL..

1_.4, was complete and prior ~o proceeding above cold shutdown.

8 Conclusions The licensee's procedures for controlling work in containment after the final closeout inspection were weak. Plant procedures did not contain guidance to *

control containment entries to perform work. after the containment cleanliness checklist, CL 1.4, was complete and prior to proceeding above cold shutdow Miscellaneous Operations Issues (92702l (Closed> Violation 50:.255/95014-01: Loss of Condensate Storage Tank Level Indication. Inspection Report UR) 50-255/95013 documented the inspectors'

review of the licensee's cold weather protective measures. That review identified previous loss of level indication on one of two level indicators to the Condensate Storage Tank (T-2) and Domestic Water Tank (T-7) due to freezing of the associated transmitter. The inspectors expressed concern that level indication for T-2 could be inoperable when needed during performance of certain emergency operating procedures. IR 50-255/95002 originally identified the problem with level indication due to freezing on T-2. The problem was discussed with operations personnel in late November, 1995 and with management on December 6, 1995.

. On December 9 I 1995 one of two level transmitters, LIA-2021, for T-2 failed high due to cold weather. Level indication to T-7 was also lost. A violation was issued *

for failure to take adequate corrective measures to prevent recurrence of the loss of level indication of T-Periodic arid -Predeter~ined Activity Control (PPAC> DMW002, "C::ondensate and Primary* Makeup Water Storage Tank Level Instruments Calibrations, II was revised..

  • to explicitly re.quire that the heat and insulation on the condensate storage tank and
primary makeup storage tank level transmitters are specially checked during PPAC

.implementation.. This item.is cl~se *

II. Maintenance M1 *

Conduct of Maintenance M 1. 1. General Comments * lnsoection Scope (62703 and 617261 The inspectors observed all or portions of the following work activities:

Work Order No:

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. 246t4736:

-24104988:

24613754:

. Replace EEO splice on SV-0861 Replacement of SOLA transformer in preferred AC bus Y-10

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Repairs to overspeed trip linkage on "A" main feed pump

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24614922:

24613931A:

2471011 38:.

Surveillance Activities

T-FC-966-03:

T-FC-966-02:

CL-1.4:

00-21 B:

R0-97: *

Repairs to P-1 B main feed pump outboard bearing oil seal Rebuild spare primary coolant pump seal for P-500 MSIV CV-0501: drill and retap hole in west stuffing box Auxiliary feedwater system post FC-966 installation intermediate and high steam pressure test Auxiliary feedwater system post FC-966 installation low steam pressure test Containment closeout walk-through lnservice test procedure auxiliary feedwater pumps

Observations and Findings. The inspectors evaluated the above activities against the FSAR and found the wor:k performed under these activities to be professional and thorough. All work*

observe_d was performed with the work package present and in. active use. The inspectors frequently observed supEJrvisors and system engineers monitoring work practic*es.. When applicable*, appropriate radiation control measures were in plac~ *

Conclusions The* inspectors* observed good procedure adherence practices; In addition, see the*

specific._observations detailed. belo M1.2 SiartuD Activities following *the Refueling Outage <RFO>

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lnsD<<[!Ction Scope *(71707 and 62703)

The in$pector_s observed. var:ious plant activities and licensee response to equipment problems associated with the startup following completion of the 1996 refueling

  • outage and subsequent forced. outage * *
  • Observations and Findings Startup Issues There were several problems identified on equipment that were safety-related or col,lld have impacted safe operation of the plant and hindered the initial" startup from t~e refu"eling qutage._ The most significant problems were:

The P-1 B main feedwater (MFW) pump outboard bearing oil seal was foun~.

running hot. A portion of the seal was found damaged and was removed. *

Also, the governor would not hold operating speed. Venting of the governor oil system was require The P-1A MFW turbine failed mechanical overspeed trip testing which required rework on the overspeed trip linkage. Also, the governor was replaced due to its inability to hold steady spee *

The new AFW discharge flow controllers oscillated during surveillance testing when set to open in the time allotted by procedure, which was derived from *design basis documentation. Engineering performed an analysis to justify meeting flow requirements of 1 50 gpm within 200 seconds. The previous test criteria contained in the FSAR was 1 50 gpm to each steam generator within 177 seconds. A 50.59 evaluation was completed satisfactoril *

The "D" 'Primary Coolant Pump (PCP) seal, which was replaced during the refueling outage, had one of the three rotating s.eals faiL as indicated by pressure oscillations associated to the failed stag *

  • on *January 5, 1997, the main generator was taken off line due to degradation of several main generator isophase flex connectors. This was similar to a degradation problem which led to a forced outage in December,.

199 *

  • The licensee initiated anothe'r plam shutdown on January 6, 1997, duet~

steam leaks on both. main steam isolation valve (MSIV) actuator steam

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leakoff plugs. * This required placing the plant in cold shutdown to perform repairs. During this shutdown, the *licensee also sched~led repairs to the "D"

  • PCP seal*, and an indepth root cause of the isophase bus connection was *

initiate *

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After identifying steam leaks on both MSIV steam leakoff plugs, the licensee.*

planned to.perform steam leak repairs; however, the repairs were unable to

.be-performed. as planned; MSIV CV-0501 did not have adequate plug thickness to support a leak repafr, and MSIV CV-0510 had an excessive steam leak which limited accessibility. The inspectors reviewed the plug thickness for MSIV CV-0501 and noted that the installed plug had been

  • previously. drilled to support a leak repair activity. This plug was required to.

be replaced during the valve maintenance during the refueling outage. The plug Wa$ removed but not replaced, and a seal weld WSS placed over the drilled hole.* This was considered an Unresolved Item (50-255/96017-04 (DRP)) pending further review by both the licensee and the NRC into the acceptability of the weld repai Refueling Outage Performance Overview The inspectors viewed the licensee's overall performance as good. Several longstanding issues documented in. Inspection Report UR) 50-255/96014 were

'~.

addressed. in this refueling outage. A negative performance trend was noted early in the refueling outage concerning control of plant and.contractor activitie Management *recognized this trend and initiated actions which successfully.

addressed these issues. The inspectors, through in-field observations and attending meetings, noted good management oversight of work activitie Other areas of improvement. in licensee performance during the refueling outage were noted:

The licensee implemented an in-mast sipping technique which detected two

. _fuel pin failures. Ultrasonic testing subsequently confirmed one of the two pin failure *

The Work Control Center minimized traffic in the control room, allowing operators *to better focus* on -the plan *

Over 1600 work orders were successfully completed, indicating good management oversight of outage activitie c. Conclusion Licensee actions appeared conservative to address the equipment problems encountered during the startup from the refueling outage. The majority of problems were rework items from the refueling outage. However, overall licensee performance during the refueling outage was viewed as good. Overall plant*

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material condition should improve.from* the efforts of this refueling outag M1.3 Develooment of a Degraded Eauioment List a. lnsoection Scope <62703 and 717071 Generic Letter (GL).91-18, "Resolution of Degraded And Nonconforming Conditions," provides guidance on resolution of degraded and nonconforming conditions affecting safety-related systems structures and components. The licensee dev_eloped a list of degraded equipment prior to plant startup from the refueling outage~ The "inspectors reviewed.the list for thoroughness and potential * *

conflict with GL 91-18 guidance or noncompliance with 10 CFR 50 Appendix B

.. "Corrective Action." Specifically, the inspectors looked for timely and adequate r~pair or engineering safety analysis of the issues identifie Observations and Findings.

The licensee's regulatory assurance group developed a degraded equipment lis from several sources which included work orders, condition reports, and

outstanding temporary modifications. Licensee -management reviewed the list for conflicts with the licensing basis, as outlined in GL 91-18. The inspectors reviewed

    • the list for its potential impact on safety* and regulatory requirements. The development and screening of the degraded equipment list by the licensee was thorough; however, the inspecto~s found some weaknesses in the resolution of some identified deficiencies:

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Three examples of small bore piping. were identified which relied on the initial operability determination from the 1994 Diagnostic Evaluation Team

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inspection until the present refueling outage, when the lines were physically restored to conform with licensing bases (Final.Safety Analysis Report (FSAR)) requirement *

Several examples of less safety-significant pipirig supports were also identified which relied on the initial operability determination from the DE *

The inspectors questioned the adequacy of the licensee's analysis of cable tray fill and cable ampacities to.verify power cables would not excessively overheat and exceed electrical code requirements specified in the FSAR.*

Licensee analysis of the four worst case cables relied on a methodology.that has not_ had NRC approval. This item was considered an Inspector Followl,lp ltem'(50-255/96017-05 (DRP)) pending Office of Nuclear Reactor Regulation (NRR) review of the licensee's methodolog The inspectors reviewed the initial operability determinations for the piping' and*

piping supports and had no concerr:is about the operability of. these item Normally, if operability was assured based upon a prompt operability determination, plant operation.could continue *whi.le an appropriate corrective action w_as implemented to* fully restore the system's *qperation or further engineering analysis determined the system met its design function. This would be an acceptable licensee response* and would satisfy the guidance in 10 CFR 50 Appendix B and as outlined iri GL 91-18. However, the g*uidance of GL 91-18 was m_or~specific in addressing piping supports. G_L 91-18 stated that piping supports should be

.restored to *FSAR criteria by the next refueling outage. *

Conclusions The licensee's attempt to identify possible equipment deficiencies, based on the guidance of.GL 91-18, was found to be.thorough. However, the inspectors identified ex~mples of weaknesses.in. timely followup on corrective action of the identified deficiencies to assure that problems were resolved either by re.pair of the affected equipment or that an adequate engineering safety analysis was performe There were no issues identified that would have potentially impacted plant startup from the refueling outag MS Miscellaneous Maintenance Issues (92902) *

<Closed) Unresolved Item 50-255/95009-01: Weaknesses Noted in tl)e Foreign Material Exclusion (FME) Program During the 1995 Refuel Outage. Most significantly, oil was inadvertently sprayed into the reactor cavity, and a plastic bag

was found Inside the high _pressure safety injection pump. The licensee developed administrative.procedure 5.09, "Maintenance Clea.nliness Standards," which appeared thorough. Additionally, the barri.ers around* the reactor cavity. Debris Free

  • Zone (DFZ) were made higher and sturdier to prevent foreign material entry into the DFZ. During the 1996 refuel outage, there were no significant:events involving

. foreign material control. This item.is closed.*

<Closed> Unresolved Item 50-255/95008-01: Crane Inadvertently Severed an Overhead Power Line Supplying Lighting to the Main Parking Lot and Guard.Hous Additionally, a similar event occurred the previous.inspection period when a* mobile crane collided with an overhead support structure. The lesson plan for site specific mobile crane operator training was reyised to include the use of spotters and an

. overhead.obstruction map. All personnel operating-mobile cranes were required to attend this training as part of their qualific~tion.. Additionally, the plant general manager is~ued an internal correspondenee stating that the following direction should become stand~rd operating procedure for m~bile cranes:

  • Schedul~ mobile.crane work during daylight h.ours; minimize nighttime hour$~. Prejob briefs shall include cautions. regarding boom to equipment:*

interferenc *

  • Travel paths shall be discussed to cle~rly. identify over.head* ciearance.

Mobile. crane spotters shall be used. *"

The inspectors ha~e* ~oted no* fUrther events inv~ivi11g damag.e.caused by. *mobile cranes. This item is close.

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Ill. Engineering E1

.Conduct of. Engineering E 1'. 1 *Environmental Qualification of Containment §plices

. lnsp~ctjon*Scooe (37551 and_ 62703).

During work on the service water system containment air cooler outlet vaive SV-0873, an electrical maintenance technician identified that the splice had bee *improperly performed and did not meet EO requirements.* The licensee reviewed other splices that were *performed during Field. Change "(FC) 6.23, which had originally installed the splice, and s~veral more splices were found to be incorre.ctly performed. The. inspectors reviewed the licens~e' s r.esponse to this* findin *

  • .. 14
      • Observations and Findings The inspectors questioned limiting the scope of inspection to splices performed during FC-623 because the cause could be related to training versus individual job scope. The licensee formed a team to review the issue and found that Rome cable had the improper splices. The scope of the inspection was widened to focus on Rome cable and involved 40 percent of the cable within containmen Additional review also determined that in 1994, an unrelated issue involving improperly performed EQ splices in containment prompted the licensee to randomly

. sample approximately 20 percent of the cables in containment. Some discrepancies were identified. The manufacturer of two of the cables could not be identified, but the scope of the earlier review was not enlarge The second review identified these two cables as Essex cable. The licensee *

concluded that due to similar manufacture, Essa~ cable was sometimes used in place of Rome cable.. An.engineering evaiuation determined that the Essex cable was quaiified for use in place of the Rome cable; however, the Essex cable was replaced because the manufacturer had not approved it for use in containment environment. The data base for locating Rome cable appeared.to be accurate during the 1996 cable revie *

Nearly 300 *splices were completed prior to startup from the refueling outage. Th inspectors observed several cable re-splices; no problems were identified.

  • C. *

Conclusions.

..

.The licensee planned and executed the resplicing well; however, weaknesses were noted in determining the initial scope of the proble.

ES

  • Miscellaneous Engineering Issues (92902 and 92702)'

(Closed> Liqensee Event Reoort 50-255/94-010: Open Convection Barrier Panels *

Result ih.Bioshield Concrete Temperatures in Excess of Technical Specification

. Limit. On Ma.rch 2, 1994, it was identified that two of the hinged horizontal panels of insulation comprising the convection barrier around the reactor vessel were not in the closed position. It was determined that the two open sections of the barrier allowed the concrete bioshield temperature to increase above limits established in the Technical Specifications. Analysis determined that there was no adverse impact to the bioshield. In April 1996, Licensee Amendment 171 moved the Shield Cooling system functional requirement from the Technical Specifications to the Final Safety Analysis Report. The failure to assure the concrete in reactor cavity does not overheat and develop excessive thermal stress is considered a violation of Technical Specifications 3.15. However, this matter is regarde~ as of minor safety significance and is being treated as a Non-Cited Violation, consistent with Section IV of the Enforcement Policy (NCV No. 50-255/96017-06). This item is close *

<Closed> Licensee Event Report 50-255/96-008: Fire Door 81 A Not Maintained*

Open Per Design Basis. Fire door 81 A, which separates the two mechanical equipment rooms (MERs), was discovered in the closed position. The design basis document CDBD) for the control room heating, ventilation and air conditioning system (CRHVAC) stated that the door had to remain open to assure that the two MERs remain pressuriz~d at all time to prevent infiltration of outside air in the CRHVAC system and affect the habitability of the control room envelope. Upon

.further engineering analysis, the CRHVAC system was determined to function as designed to maintain the control room envelope pressurized in accordance with the design basis, with the fire door in the open or closed position. On May 17, 1996, a surveillance was performed and pressure readings recorded in the CRHVAC MER with door 81 A closed and the system in the emergency mode.. The inspectors observed portions of the surveillance. The data confirmed that the CRHVAC would maintain a positive pressure with either CRHVAC train in service in the emergency

. mode with fire door 81 A closed. The surveillance test procedure was revised to ensure ~oor 81 A will re.main closed in the normal and emergency operating mode * * Also, there is a DBD change request submitted. This item is closed:

IV~ Plant Support R 1 Radiological P~otection R1.1 Maintenance Outapes and Daily Radiological Work Practices

. liisoection ScoD1r<71750 and-83750>

The inspectors observed radiological worker activities during the various applicable outages* detailed in this inspection report, and also monitored radiological practices during daily plant tour Observations and Findings *

During the applicable maintenance outages radiation technicians were visible at the job sites. The technicians *took appropriate actions and surveys in accordance with good ALARA practice *

  • . Conclusions RS
  • The _inspectors concluded that radiological practices obs~rved d~ring the maintenance outages and plant daily walkdowns were adequate. The inspectors had *no concern Miscellaneous Plant Support Issues

<Closed> LER 50-255/96-001: Failure to Test Duplicate Equipment. Technical Specification 3.3.2.f requires, in part, that prior to initiating repairs, all valves and interlocks in the system that provide the duplicate function shall be tested to

-.

.J

.. i determine operability. Contrary to the above, on-January 3, 1996, air header blowdown valve (MV-CA522) was repaired without testing the alternate train Containment Sump and Safety Injection and Refueling Water Tank outlet valve The *licensee documented three reasons for failure to test the valves:

1.

The outlet valves had passed previous operability surveillance.

The repairs could be completed within the TS 3.3.2.f LCO time, but the pre-repair testing could no ~-

A personnel safety hazard existed b.ecause a test rig was serving as the air system pressure boundary, and it would have to remain in place umil the*

pre-repair testing was completed and the affected air system depressuriz~ Additionally, the licensee determined that while* it was possible to place the_plant in cold shutdown in order to perform the testing and repair, it was imprudent to put the plant through a cooldown/heatup transient in order to accomplish this activit The licensee submitted Technical Specifications change request on February 6, 1996, and Amendment 172 was approved on September 26, 1996, to remove the requirement for cross-train testing. This licensee-identified and corrected violation of Technical Specifications 3.3.2.f.is being treated as a Non-Cited Violation (.50-255/96017-07), ~onsistent with Section Vll.B.1 of the NRC Enforcement Polic This L~R is close~.

V. Management Meetings

..

.X1 Exit *Meeting Summary The. inspectors *presented the* inspection results to members of licensee

..,

management at the conclusion of the inspection on *January 15,.1997. No

_proprietary information was identifie * -

PARTIAL LIST OF PERSONS CONTACTED Licensee R. A. Fenech, Vice President, Nuclear Operations

-

T. J. Palmisano, Plant General Manager K. P. Powers, Nuclear Services General Manager G. B. Szczotka, Nuclear Performance Assessment Manager R. J. Gerling, Design Engineering Manager T. C. Bordine, Licensing Manager D. W.. Rogers, Operations Manager J. P. Pomeranski, Maintenance and Construction Manager D. P. Fadel, System Engineering Manager

D. G. Malone, Shift Operations Supervisor M. P. Banks, Chemical & Radiation. Protection Services Manager K. M. Haas, Training Manager.

S. Y. Wawro, Planning & Scheduling Manager M. E. Parker, Senior Resident Inspector, Palisades P: F. Prescott, Resident Inspector, Palisades

IP 37551:

IP 61726:

IP 62703: _

IP 71707:

IP 71750:

IP 83750:

_IP 92702:

'1p 92902:

INSPECTION PROCEDURES USED Onsite Engineering Surveillance Observations Maintenance-Observation Plant Operations Plant Support

-_Occupational Radiation Exposure

_ Followup oil Corrective Actions for Violations and Deviations

  • F_ollowup - Maintenance ITEMS OPENED 50-255/960.17-01 VIC Failure to maintain PCS temperature above the Tectu1ical Specificati_on limit of 525° ~hen placing turbine on-line -

_Failure to maintain Radial Peaking Factors within Technical

  • * *

Specification limit 50-255/96017-04 UR Non~code :repair *to main steam isolation. valve.

'

.

    • .

.

-.

.

.

50-255/96017-05 IFI * * Review of licensee;s methodology for analyzing *cable ampa~ities and cable fray fill *

- ITEMS CLOSED.

50-255/95014-01 VIC. * Loss of cond_ensate storage tank level indi~ation: ** -*

50-255/95009-01 URI * Weaknesses noted. in :the foreign material exclusion (FME)

program during the 1995-refuel outage.*

.

-.

.

.

,

50-255/95008-01 URI Crane inadvertently severed an o~erhead po-wer line supplying:

  • 50-255/94-010 50-255/96-008

. lighting to the main. parki_ng lof and guardho~se.

-LER Open convection barri~r panels result in bioshieid concrete temperatures in excess of Technical Specificati9n limit

-

-

.

LE~. 'Fire door 81 A not maintained open per design basis 50-255/96-001

_. --LER-Failure to test duplicate equipment 50-255/96017-06 NCV.Open convection barrier panels resuit in bioshield conc:rete

  • temperatures in-excess of TS limits
  • 50-255/96017-07 NCV _ Failure to test duplicate equipment *
  • -

.. :

. DET DFZ.*

DRP

    • Ea FC

. FME FSAR GL GOP GPM IR LER LCO MER MF MSIV NCV

  • NRC

.. NRR

.PDR PIDAL PPAC PPC PVC RFO SS sv TS LIST OF ACRONYMS USED As Low As Reasonably Achievable Auxiliary Feed Water *

Core Exit Thermocouple Code of Federal Regulations Check List Control Room Heating Ventilation & Air Conditioning Control Valve Design Basis Document Diagnostic Evaluation Team Debris Free Zone Division of Reactor Projects

  • General Operation Pro~edur Gallons Per Minute Inspection Report:.

.-Licensee Event Report Limiting Condition of Operation Mechanical Equipment Room Main Feed Water

.Main Steam Isolation Va!ve

-.

.Non-Cited Violation:

..

.

  • Nuclear Regulatory Commissio Office of Nuclear Reactor Regulation
  • Primary Coolant Pump Primary Coolant SystelT Public Document Room Palisades lncore Detector.Algorithm
  • Periodic & Pr~determined Activity Control Plant Proc~ss COITIPUter

-.. :Refueling Outage

~ystems Structures & _Components.

_Solenoid Valve Technical Specification *

...

...

QCOP QCOS QGA RCIC RCU RFP RG RHR *

RWCU SBGT SBLC

.. S **

SLRC SSM TD TS UFSAR URI Vl_O VOTES Quad Cities Operating Procedure Quad Cities Operating Surveillance Procedure

' Quad Cities General Abnormal Procedure Reactor Core Isolation Cooling Syste Refrigeration Condensing Unit Reactor Feed Pump Regulatory Guide Residual Heat Removal Reactor Water Cleanup Standby Gas Treatment Standby Liquid Control

. Safety Evaluation Report Steam Line Resonance Safe Shutdown Makeup Time Delay

  • Unresolved l~em Violation Valve Operation Test and Evaluation System 30
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