IR 05000155/1997010

Insp Rept 50-155/97-10 on 970802-1007.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support

ML20202C950
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/26/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20202C931	List: IR 05000155/1997010 ML20202C927
References
50-155-97-10, NUDOCS 9712040106
Download: ML20202C950 (20)

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U.S. NUCLEAR REGULATORY COMMISSION l REGION 111 I

Docket No.: 50 155  ;

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License No.: DPR-06

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Report No.: 50-155/97010(DRP)  !

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, Licensee' Consumers Energy

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i Facility: Big Rock Point Nuclear Power Plant

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Location: 10269 U.S. 31 North i

Charlevoix, MI 49720 t

Dates: August 2 October 7,1997

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inspectors: R. J. Leemon, Senior Resident Inspector -

C. E. Brown, Resident inspector -

E. R, Schweibinz, Project Engineer Bruce L Burgess, Chief I Approved by:

Reactor Projects Branch 6

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EXECUTIVE SUMMARY Big Rock Nuclear Power Plant NRC inspection Report No. 50155/97010 This routine inspection covered aspects of licensee operations, engineering, maintenance, and plant support.

• Overall, the licensee maintained a high level of piofessionalism through the final phases of operation, shutdown, and defueling. The operators remained attentive to their duties and all scheduled surveillance tests and prt,ventive maintenance tasks were performed as scheduled.

Operations e The inspectors concluded that the overall final shutdown was performed in a controlled

.aanner. (Section 01.2),

e No unnecessary dirtractions were allowed in the control room, despite the potential distractions from the licensee's plans associated with the permanent shutdown of Big Rock Point. (Section 01.2)

Maintenance o Maintenance and surveillance activities were correctly performed and accurately documented. (Section M1.1)

e The inspectors concluded that the work to inspect and repair the No.1 spent fuel pool heat exchanger was well performed Thorough planning and extensive supervisory *

interaction resulted in the work being completed with a dose expenditure that was below the total projected person dose and without any contaminations. The time that the N spent fuel pool heat exchanger was removed from service also was reduced due to effective planning and supervisory involvement. (Section M21)

e The inspectors concluded that the licensee's decision to maintain redundancy by keeping both emergency heat exchanger bundles in service when Technical Specifica6ons allowed isolating one was conservative. (Section M2.2)

e The inspectors concluded that operating with an elevated emergency condenser temperature was not a safety concern after reviewing the Final Hazards Summary Report (FHSR) safety analysis for the emergency condenser and ATWS (anticipated transient without scram) events. (Section M2.2)

e A work planner demonstrated a good questioning attitude and attention to detail by noting <

the absence of FHSR requirements in the fuel-transfer cask preparation procedur (Section E3.1)

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Enaineenna l e The inspectors concluded that the engineering staff performed wellin correcting a possible discrepancy between the procedural and FHSR requirements for the fuel transfer  !

safety brake tag line clearances. (Section E3.1) ,

o Engineering anel"ses supporting defineling moves were either incomplete or inaccurat ,

The fuel bundle; cedure was properly modified only after the inspectors identified i concems. (Sectic E4.1) l

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e The overall final defueling operations were completed in a methodical, safe, and cautiour manne ,

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e The inspectors noted a relaxed. safety focus as indicated by the 'icensee accepting tw>

poor engineering assessments, init; ally directing that a heavy load be left unattended over the reactor, accepting a verbal procedure change, and treating dynamometer indications as spurious fluctuations. Overall, despite the flawed analyses and the relaxed safety .

focus, the licensee safety defueled the reactor in a cautious and methodical manner with !

no fuel damag Plant Support e On one occasion, radiation protection technicians did not observe a worker lying down in a contaminated area until alerted by the inspectors. However, the technicians' overall ,

efforts were responsible for the total person-dose being under the projected dose  :

estimates. In addition, no personnel contaminations occurred during the inspection  ;

period. (Section R1.1) ,

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Reoort Details  !

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B!mmaryJB!anlEth!1 1, Operations  !

i The licensee operated the plant in coastdown, with all control rods fully withdrawn, from the beginning of the inspection period until the final plant shutdown on August 29,1997. As planned, i the operators tripped the turbine at 10:32 a.m. (EDT), and manually scrammed the reactor (inser1ed all control rods) at 10:33 a.m., ending 35 years of plant operation. A final core off load began or. September 5 and continued until 5:56 p.m., September 20, when the last fuel bundle was removed ' rom the reactor and placed in the spent fuel pool. Once all of the spent fuel (onsite) wa3 stored in the spent fuel pool, the licensee began the decommissioning phase of the plant life. B,e letter dated September 23,1997, the licensee certified the permanent cessation of .

operation (s) snd permanent fuelremovalfor the Big Rock Point Nuclear Plan F

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01 Conduct of Operations

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0 General Comments (71707)

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Using inspection P ocedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. Specific events and findings are detailed in the sections below.

01.2 Einameas19tR19Ld3Lw]

The inspectors observed the final reactor shutdown activities in the control room on August 29.- The shutdown was characterized by clear operator communications, attentive reactor engineer oversight, and effective shift supervision entrol. Shift management had completed extensive and comprehensive pre evolution briefings before starting the shutdown. In addidon, to ensurs all personnel understood what was expected, the shift supervisor (SS) held mini-briefs immed!ately before moving into each phase of the

- shutdown. The inspectors noted that the appropriate procedures and control-rod pull sheets were in active use. A shift turnover, at low power, was well planned and executed. The operators focused closely on plant operations and controlled the plant shutdown and cooldown without being distracted by the extensive licensee ceremonies for the final reactor shutdown. Licensee management did not allow any unnecessary personnel or other distractions in the control room. The inspectors concluded that the overall shutdown was conducted in a controlled manne .3 lings _ ele _eping During frequent tours of the containment and other buildings, the inspectors noted that trash and debris did not accumulate. The general cleanliness remained good throughout the plan Miscellaneous Operations issues (92700)

As a minimum, the inspectors reviewed the existing open items under l'te folios n criteria: (1) the issue is not applicable to a permanently shutdown or decommissionmg

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reactor, (2) the issue does not raise potentially generic concerns, (3) the issue does not l

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involve any pending enforcement action or open investigation, and (4) the issue does not l j involve any indication of willful violation ;

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08.1 The licensee's corrective actions for the items listed below were of concem only for j _ power operations and are therefore administratively closed:

[Qlated) Follow-Up Item 50155/95011-03: use and control of caution card l

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(Cip1td)f_pflQtVp.ltem 50155/96003 92: final hazard summary report .

clarification of spent fuel poolissue l

. (Closed) Follow Uo item 50155/96004 04: concerns over use of contaminated

laundry as a store room.

!- (CloAAdlynrgnpired item 59155/96006-0j: weak turnover during shifts j (adequacy of shift manning and shift-turnover). ,

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(Closed) Follow-Up ltem 50-155/96008-01: No. 2 reactor protective system motor

generator over voltage trip . review the root cause evaluatio {

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(Glqsed) Follow 3p item 50155/901008-03: heavy component storage inside containmen .

(Closed) Violation 50-155/96010-01: motor operated valve (MO 7062) closed

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without SOP-6 (standard operatir,g procedure) actions performed, j

[CJple_d1Molation 50155/97008-01: control rod drives left with high insertion 1

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speeds after time testin ,

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(Closeal Violation 50-155/97008-02: contair nent equipment lock inner door left  !

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08.2 Using Inspection Procedure 90712, the inspectors performed an in-office review of licensee event reports (LERs) at the time of issue. Further inspection efforts (onsite) are no longer warranted due to the licensee's permanent fueiremoval. The following LERs are closed:

(Closed) LER 50-155/95006: inoperable emergency-diesel generator and J

standby-diesel generato (GloandRiiRfD.155M0_01: automatic reactor scram during plant startup.

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l (Gl91L d lib.5035h@60J3: automatic reactor scram initiated by turbine tri .3 13s99s related to 1991ligalated Enforcement EA 95-057:

The licensee's performance in inadequately repairing, testing, and operating fire system

- dual basket strainer (BS-5761) was discussed in detail in inspection report (IR) 50155/95006. Escalated enforcement (EA 95-057) and extensive licensee

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corrective actions in 1995, resulted from IR 95006. The licensee's corrective actions addressed the specific procedural problems, training deficiencies, and staff practices that contributed to the inadequate work on BS-5761. Based on the inspectors' reviews of subsequent maintenance on and operation of BS-5761 and observed licensee performance, the inspectors concluded t! 3t the corrective actions taken were effective in reducing procedural adherence problems by operations and maintenance department personnel, particularly for work activities associated with safety related components. The items listed below are closed:

(Closed) Violation 50-155/EA 95-057-01013: violation of Technical Specification 11.3.1.4.D for approximately 78 day (Closed) Violation 50-155/EA 95-057-01023: inadequate training of personnel concerning 'he O-List [ Quality List).

[Clo s edLMqlajlo rL50dM/g A_95-057-01033: inadequate u.structions for repair of BS 576 (Closed) Violation 50155/EA 95-057-01043: lack of procedure for operating B S-576 , .

(ClosedLWojal!oAMdMLE6 95-057-010M: failure to inspect work associated with BS-5761 repai (Closed) Violation 50155/EA 95-057-01063: failure to perform post maintenance test on BS-576 (CI.osedjXiolajign 50155/EAJ5-0_5101073 inadequate corrective actions for numerous DP [ differential pressure) alarms on BS-5761 during test II. Maintenance M1 Conduct of Maintenance M1.1 General Comments Inspection Scope (62703) (61726)

Using inspection Procedures 62703 and 61726, the inspectors observed all or portions of the following work activities:

Maintenance Activities

• WO CLP 12710864: change trip setting of transfer cask to 1200 pounds

• WO CLP-12611833: fabricate stirrup for transfer cask

• WO CLP 12710868: remove stirrup cable from transfer cask

• WO EPS 12710801: breaker trip when stand-by EDG [ emergency diesel generator) closed onto 28-bus l 6 l

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e WO SFP 12710771: feel pit heat exchanger N * WO FHS-12710715: support return of new fuel SufXeillan;eAtlyllies

• TR 02: refueling interlocks test e TR 07: fuel handling equipment safety check

• T7 04: reactor protection logic test

• T7 20: diesel fire pump auto start test

• T7 28: emergency diesel generator auto-start test

• T60-03: liquid poison system firing circuit test

• MFHS 1: transfer cask preparation for fuel movement Qb_temations and Findinas The inspectors monitored the maintenance and surveillance activities closely au.. a h period leading to the final plant shutdown on August 29. The licensee evaluated all c htanding work orders for cancellation due to the imminent plant shutdown. The evaluations included personnel dose estimates, effect on safety, and Technical Specification (TS) requirements before the licensee made a decision on a proposed cancellation. The inspectors' reviews of the licensee's actions did not reveal any inappropriately canceled maintenance. All scheduled surveillance tests and maintenance tasks (preventive or emergent) to maintain the plant operating ner TS requirements were completed as scheduled. All observed work was thoroughly completed and properly documente c. Conclusions Maintenance and surveillance activities were correctly performed and accurately documented. The inspectors did not observe any decrease in performance or improper work or surveillance postponement due to the imminent final plant shutdown.

M2 Maintenance and Material Condition of Facilities and Equipment M No.1 Spen 1Euel Poo11SFPWeat Exchanger Tube inspection Lnspection Scopp From August 18 through 22,1997, the inspectors observed portions of the work to inspect and repair the No.1 SFP heat exchanger (HX). The inspectors also reviewed the results of tests on the HX tube Qbservations and Findings To ensure reliable heat removal capability from the SFP after the final shutdown and durin;, decommissioning, the licensee had worked on the No. 2 SFP HX, including eddy current inspection of the HX tubes. In response to having to plug 13 tubes in the No. 2 SFP HX (IR 50-155/95009, Section E2.1), the licensee expanded the work scope to include the No.1 SFP HX even though there was no externalindication of tube leakag . _ . . _ . _ _ _

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The inspectors attended the pre job briefing for the work. The operations, maintenance, and reactor protection supervisors gave extensive and detailed presentations which included lessons teamed from previous work on SFP HXs, possible leakage through isolation valves, personnel dose, contamination, and limiting the time that the SFP HXs would be out of servic While monitoring the werk, the inspectors noted that the work site had been v' ell prepared, including limiting the amount of radiological draping to the minimum amount necessary to control the possible spread of contamination; this was important as the work was inside the protected containment floodable volume. Additionally, the planners had sequenced the work to ensure the workers would spend the minimum time in the immediate area of the SFP HXs, reducing the total person dose. The work sequence had been changed from the sequence used on the No. 2 HX so that eddy current testing was completed on all the HX tubes before doing leak testing. The eddy current testing results revealed three tubes with greater than 60 percent weil thickness loss, but no through wall holes. However, the subsequent prcssure test revealed four leaking tubes. The inspectors noted that a condition report was generated and that the evaluation stated that the eddy current testing method used was not capable of detecting leaks in the tube sheet or tube support areas. This was the reason that the licensee had not relied solely on the eddy current testing results. The inspectors also noted that the licensee had reviewed the results of the testing thoroughly before plugging the defective tubes and closing the heat exchanger, in contrast to the sequence of work on the No. 2 SFP HX (IR 97009, Section E2.1). The inspectors had no further cancerns with the testing result Overall, the licensee had performed work effectively without any personnel contaminations at a dose expenditure that was below the projected dose, and according to the schedule developed from the procedure. The inspectors also noted the frequent presence of supervisory personnel at the lob site. The licensee minimized the time that the No.1 SFP HX was not available for us Conclusions The inspectors concluded that the work to inspect and repair the No.1 spent fuel pool i heat exchanger was well performed. Thorough planning and extensive supervisory interaction resulted in the work being completed with a dose expenditure that was below the total projected person-dose and without any contaminations. The time that the N spent fuel pool heat exchanger was removed from service was reduced due to effective planning and supervisory involvemen M2.2 Eleyated Emergeacy Heat Exchanaer Temperature The inspectors evaluated the licensee's actions in response to increasing emergency heat exchanger temperature. The operators had noted a slowly increasing secondary side (shell) temperature after cycling the emergency condenser outlet valve (MO-7063)

for timing with surveillance test T90-26, * Emergency Condenser Valve Operability Test?

on July 19,1997. In an attempt to improve the leak tightness of the valve, the operators cycled MO-7063 again on July 23,1997, with no improvement. The temperature increased about a degree a day ar.d finally stabilized at 137*F in early August. Technical I

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Specification requirements allowed the licensee to isolate one tube bundle in the i

< emergency condenser, but the licensee had chosen to maintain redundancy by keeping

both bundles in servic The inspectors reviewed the safety analysis in FHSR Section 15.8, " Anticipated Transients Without Scram," Revision 2, and determined that the accident analysis took no credit for emergency condenser operation. Therefore, continuing to operate with the -  :

emergency condenser shell at 137'F was covered by the safety analysis. The inspectors i

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independently verified the shell side temperature and temperature trend by morJoring the tube bundles' telltale temperature recorder. When the plant shutdown on August 29,  !

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1997, the shell side temperature was still stable at 6 out 137" Conclusions . ,

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After reviewing the FHSR safety analysis for the emergency condenser and ATWS

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(anticipated transient without scram) events, the inspectors concluded that an elevated y '

emergency condenser temperature was not a safety concern. The licensee chose to track the temperature and maintained redundancy by keeping both tube bundles in service although Technical Specifications allowed one bundle to be secured.

I M8 Miscellaneous Maintenance issues (92902)

As a minimum, the inspectors reviewed the existing open it sms under the following  ;

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criteria: (1) the issue is not applicable to a permanently shutdown or decommissioning reactor, (2) the issue does not raise potentially generic concerns, (3) the issue does not involve any pending enforcement action or open investigation, and, (4) the issue does not involve any indication of willful violation M8.1 The licensee's corrective actions Gr the items listed below were of concern only for

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power operations and are therefore administratively closed:

(Closed) Follow-Up_(tem 50-155/95008-0]; multiple concerns with spent fuel-pool heat exchanger cleaning job relative to a design basis accident with conservative heat load [ Closed) Follow-Up item 50-155/9501102: effectiveness of work control  !

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measures for fire protection expansion tan i (Closed) Unresolved item 50-155/96002 03: lack of evaluation of breakdown in work control process - wrong pipe hanger cu !

LQ.l_osed) Follow-Up Item 50155/96002-04: review of breaker testing procedures and data recording requirements - molded case circuit breaker test issues (Closed) Unresolved item 50-155/96002 09: breakdown in work control on MO-707 ,

LCjottd) Follow-Up Item 50-15}/96005-02: inspectors evaluation of reactor-depressurization system (RDS) battery failur .

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(Closed) Violation 50-155/96006-02: inadequate post maintenance testing on RDS *B".

{Glqsed) Violation 50155/96010-03: failure to perform surveillanc (C191gsflyigtahon 50-155/96Q10-M: fire barrier non functiona {Q191ed) Violation 50-155/96012-02: two examples of violations of 10 CFR 50, Appendix B, Criterion (Closed) Follow Up Item 50-155/97002 05: material condition concerns with controt rod drive syste (Clq}1d. ) Unresotypd item 50155/97004 02: protective devices installed on DC breakers different than the original desig (Closed) Follow Uo item 50-155/97007-01: corrective action to be reviewed for emergency lighting unit battery problem M8.2 Using Inspection Procedure 90712, the inspectors performed an in office review of licensee event reports (LERs) at the time of issue. Further inspection efforts (onsite) are no longer warranted due to the licensee's permanent /ue/ removal. The following LERs are closed:

(Closed) LER 50-155/95008: a Technical Specification surveillance requireme exceede LQlo_ sed)J,F,R 50-155/96012: diesel fuel 92 day Technical Specification surveillance requirement inadvertently surpasse Ill. Enaineerina E3 Engineering Procedures and Documentation E Safety Brake Trio-Line Clearance {0ffection Scope (37551)

The inspectors reviewed the engineering staffs involvement in developinn corrective actions for a noted deficiency in procedure MFHS-1," Transfer Cask P- tration For Fuel Movement," Revision 2 Observations and Findinas On August 25, while doing a review of the FHSR (final hazards summary report) for an unrelated change, a work planner noted that MHFS 1 did not include the FHSR dimensions for the safety brake trip-line adjustment for the transfer cask. The FHSR specified that clearance (slack) should be set at less than 0.37 inches, The existing

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c procedure only stated, " adjust trip line to remove slack without tension." The work planner discussed the concern with the maintenance engineer. The engineer initiated a condition repor1(C-BRP 97-0455) and the licensee delayed the start of defueling for 3 days to resolve the issue. When interviewed, the mechanics estimated that the actual 3 clearance could have been as much as 1.625 inches.

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The inspectors noted that the condition report received a high priority for resolution; a team was formed and a detailed investigation started immediately. The inspectors'

reviews of the FHSR, Section 9.1.4," Fuel Handling System," and Sections 15.7.1.2.1 and 15.7.1.3.1, " Fuel Transfer Cask Drop," along with a review of the O-List, revealed that the fuel transfer cask ir not a safety-related component and that fuel handling inside the reactor or the spent fuel pool (SFP)is not a safety related actio The licensee's concern with the configuration of the safety brake trip line related to over stressing the refueling deck overhead crane in the event of a cask-drop accident. The cask weighs about 24 tons and the crane is rated at 75 tons. The maintenance engineers were concernea that too much safety brake trip-line slack would allow too much momentum to be developed before the safety brake engaged in the ever, of a cask dro . Conversely, the operators have to rotate the transfer cask while transferring fuel and could trip the safety brake if there was not sufficient clearance if the safety brake was tripped with the cask in a position where a fuel bundle could not be safely lowered into the reactor or the SFP, the operators would have had to provide cooling to the cask for about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while a maintenance crew reset the safety brake. Providing cocling to a fuel bundle in the transfer cask is a safety related action. The licensee's investigation revealed that the safety brake trip-line could be, and had been, set within the FHSR tolerance by the existing procedure. Actually measuring the slack was difficult, but setting the slack per the procedure was not. The inspectors otserved the operators do a complete dry-run walk through of the fuel moves, including measuring how many degrees the cask could be rotated without tripping the safety brake. Several members of the engineering staff, including the maintenance and system engineers, were directly involved

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in resolving this deficiency, including being present defing all measurements, adjustments, and walkthrough Conclusions The inspectors concluded that the licensee's staff performed well in identifying and correcting a possible discrepancy between procedural and FHSR requirements pertaining to the fuel transter cask's safety brake trip-line clearances. A work planner demonstrated a good questioning attitude by noting the discrepancy, and the engineering staff aggressivery pursued resolving the issu E4 Engineering Staff Knowledge and Performance E Enaineerina Support for Defuehna Coerations inspection Scope From September 5 through 20, licensee staff permanently removed all fuel from the reactor. The inspectors verified that procedure O-FFl-1," Fuel Bundle Removal Procedure," Revision 16, agreti with the descFption in the FHSR and obse,ved the

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t defueling operations, ncluding attending meetings. For the most part, the defueling went smoothly, according to procedure and licensee expectations. However, using normal force, the operaters could not extrael 5 of the 84 fuel bundles from the core. The 4 inspectors evaluated the licensee's actions, pcrticularly engineering assessments, supporting defuelin b, Qbservations and Findinos Procedure O-FFI 1, Section 2.0," PRECAUTIONS AND LIMITATIOt% 7 wagraph ,

states, in part, that, " Technical justification is required to use grapp e m ch have pulled

>1200 lbs." During defueling, the inspectors witnessed the operators aewd 1200 pounds on a fuel bundle on three occasions. The first occasion was when operators were " shaking" tiie lifting cable in an attempt to free a bundle after the transfer cask hoist had tripped off at about 820 pounds-pull. The shaking resulted in the dynamometer telltale showing an increase of 450 pounds. The second occasion was while using a pry bar and assist cable to increase the lifting force above 900 pounds (allowed by O-FFI-1 after receiving management permission). The third occasion was while attempting to pull 16-19 (one of the last five bundles) out of itr fuel channel. The pry bar slipped to a corner of a triangular stirrup, causing a jolt, which resulted in the dynamometer telltale going to 1300 pounds, the assist cable breaking, and the operator almost falling into the flooded

.. reactor vesse In response to a condition report on exceeding 1200 pounds the first time, a reactor engineer (RE) reevaluated three fuel bundle lifting components: the grapple, the lifting cable, and the fuel bundle. The engineer concluded that the actualiifting limit was 1500 pounds (the maximum grapple strength) and that the grapples could be returned to service after a visual examination for deformation. However, the RE's evaluation was non conservative b that it did not address why the procedure limited the lifting force to 1200 pounds and it improperly focueed on ti,e maximum force that could be applied rather than just on whether the grapple wa. suitable for re-use. The inspectors verified that the operators had examined and installed a spare grapple before restart:ng fuel moves after each occasion of exceeding 1200 pound When the accidental 1300 pound pull failed to extract 16-19, the on-sn.1 management group --shift outage manager (SOM), shift superintendent (SS), and refueling senior reactor operator (SRO) - discussed the stuck bundle condition and initially directed the operators to leave the bundle grappled with the transfer cask suspended over the reactor and to exit containment while a course of action was developed. However, when the inspectors expressed concern about what control measures would be in place while an unattended " heavy load," the 24-ton transfer cask, remained suspended over the reactor core, the shift managers reconsidered their decision and had the operators ungrapple the fuel bundle and move the Mnsfer cask to the storage cradl Later, the inspectors heard the RE tell the SOM that, based on the engineering

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assessment for the lifting components, "the operators could go up to 1500 pounds lifting force." From logs, the inspectors found that the SS had already given verbal permission to exceed 1200 pounds to the refueling SRO before the assist-cable broke. The inspectors asked how the operators could purposely exceed 1200 pounds pull on a bundle without a procedure revision when tec!,uical justification to re-use grapples that I

had exceeded 1200 pounds pull was still required. After reconsideration. the SOM and i

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IR ovv d ihA a iwWviv )weem arqurwt tefere the operators could procee /\fiO!kihd IW tv qui nowg in, tirtog fette to 1500 pounds, the licensee al6%ivi lh4 an 9414a tolushvo hM twemmended replacing the pulley axles on the WMmWovlw nuJ the heder utA 4! ING pounds-force had been exceeded. Also, the 1991 uanufctum thd not wnwr with vseceding 1200 pounds pull on irradiated fuel Iqin@on Indonuv hpononc, had shown that irradiated fuel bundle tie-rods could be unW9% INaud 904 vid bundles could fait at a much lower stress value than new ly an tio n, lhe laconnu rovaed 0-FFl-1 to set the electrical-lifting force at 1200 pounds in mi eliumpt to romWe the last five bundles. The inspectors questioned whether "other umanC of fluoing the bundles would be allowed, specifically using the pry bar and assist-Geblo 41 shaking the lifting cable. The licensee agreed with the inspectors' concern and prohhied uting any method other than a straight lift. After removing the assist cable fsom the tranbler cask winch and resetting the lifting limit, the operators successfully lidabforred throo inore bundle At a plant review committee (PRC) meeting, the RE proposed shaking the lifting cable, aher establishing about 1200 pounds pull, in an attempt to free the two stuck bundle Ibe recommendation was based on a civil engineer's analysis that showed the added kace due to a 3 foot horizontal deflection would be less than 10 pounds. A second civil engineer had independently verified the calculation using the same modeling techniqu *

T he engineer's evaluation assumed that the vertical component of the total force would be a constant value and stated that small cable length changes, required by shaking, would b2 provided by the cable either compressing or shifting in the pulleys. Following estensive discussion, the PRC dismissed the previously noted 450 pound dynamometer deflection as an inaccurate, transient needle fluctuation and accepted the recommendation. After the PRC meeting, the inspectors discussed their concern (that the model used to analyze the lift was incorrctt) with the plant manager. The inspectors'

analysis showed that changes in length could only be accomplished by eithe t' -

dynamometer or the cable acting as a spring (stretching) therefore the change in force would be much higher. The plant manager disagreed with the inspectors' analysis but agreed that caution was needed. The PRC subsequently approved steps to initially establish a lower value vertical pull (less than 1100 pounds), pull horizontally slowly while monitoring the change in the dynamometer reading, and stop alllifting attempts if the change in force exceeded 10 pounds. The actual results showed an increase of about 100 pounds with about a 6-inch horizontal pull. The fuel manufacturer subsequently disassembled and reassembled the two stuck bundles inside the reactor. The operators then transferred the bundles to the SFP without t.ny further difficult The inspectors determined that no violations occurred. As noted in Section E3.1 above, the fuel transfer cask and fuel moves within the reactor or SFP are not classified as

" safety-related." At the exit meeting, in response to the inspectors inquiry if the two examoles of poor engineering assessments represented overall engineering performance, the plant manager stated that the two assessments had been inadequate, but he did not feel that they represented overall engineering performance. Additionally, the plant manager felt that the engineering staff was heavily involved with overall operation .. . . - . - - _ .- - .- --

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- (Closed) LER 50-155/97004: potentialloss of DC power to primary containment spray and liquid poison system IV Plant Support

- R1 - Radiological Protection and Chemistry Controls (71750)

R General Comments _ jnsp1ction Scope Using Inspection Procedures 71707 and 71750, the inspectors made frequent tours of the radiologically protected area (RPA) and discussed specific radiological controls with the

- ALARA (as low as reasonably achievable) coordinator and various radiation protection (RP) technicians. The inspectors observed plant conditions and licensee RP practice Observations and Findinos The inspectors noted involvement by RP personnel in planning and executing work, as reflected in the preparation for the final plant shutdown and decommissioning. As low as reasonably achievable (ALARA) considerations were primary factors in the licensee's decision to decontaminate room 444 (inside containment), allowing storage of the reactor head in a remote location and thus reducing personnel dose during the decommissioning phase. The inspectors also observed that shoring was used to support the reactor head in room 444, instead of the reactor-head stand, which allowed shielding to be placed around the reactor head to further reduce the dose rate. Storing the reactor head in room

444 also allowed moving the re9ctor-head thermal shield to the main containment floor, a location that would simplify early disposal of the thermal shield. The thermal shield was

- bagged in a large, specially-manufactured bag to reduce the possible spread of contamination, a practice developed from lessons learned during previous thermal shield move The inspectors noted that RP technicians were working closely with the operations and maintenance personnel during performance of evolutions in contaminated or high dose areas, especially during defueling operations. The technicians provided close RP oversight including instructing personnel to move to low-dose areas if a job was delayed, and monitenng contamination control practices. The inspectors observed RP personnel giving highly detailed presentations at each pre job briefing to ensure that all personnel were thoroughly knowledgeable of the radiological conditions, precautions, and lessons learned from previous evolutions. The inspectors noted that the high level of involvement by RP personnel resulted in the scheduled work being done under the projected personnel dose estimates and without any personnel contamination events. However, while observed individual RP personnel performance was good, RP practices by others

- were not as good. During preparations to de-tension the reactor head, the inspectors observed and pointed out to the RP technicians that a worker was lying down in n contaminated area (discussed fully in IR 50-155/97011).

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RE agreed that a procedure revision was required before the operators could procee After further review, to support revising the lifting force to 1500 pounds, the licensee identified that an earlier evaluation had recommended replacing the pulley axles on the dynamometer and the transfer cask if 1200 pounds-force had been exceeded. Also, the fuel manufacturer did not concur with exceeding 1200 pounds pull on irradiated fuel bundles. Industry experience had shown that irradiated fuel bundle tie-rods could be unevenly loaded and old bundles could fail at a mmh lower stress value than new bundles. The licensee revised 0-FFI 1 to set the electrical-hfting force at 1200 pounds in an attempt to remove the last five bundles. The inspectors questioned whether "other means" of freeing the bundles would be allowed, specifically using the pry bar and assist-cable or shaking the lifting cable. The licensee agreed with the inspectors' concern and prohibited using any method other than a straight lift. After removing the assist-cable from the transfer cask winch and resetting the lifting limit, the operators successfully transferrad three more bundle At a plant review committee (PRC) meeting, the RE proposed shaking the lifting cable, after establishing about 1200 pounas pull, in an attempt to free the two stuck bundle The recommendation was based on a civil engineer's analysis that showed the added force due to a 3-foot horizontal deflection would be less than 10 pounds. A second civil engineer had independently verified the calculation using the same modeling techniqu The engineer's evaluation assumed that the vertical component of the total force would be a constant value and stated that small cable length changes, requirer' by shaking, would be provided by the cable either compressing or shifting in the pulleys. Following extensive discussion, the PRC dismissed the previously noted 450 pound dynamometer deflection as an inaccurate, transient needle fluctuation and accepted the recommendation. After the PRC meeting, the inspectors discussed their concern (that the model used to analyze the lift was incorrect) with the plant manager. The inspectors'

analysis showed that changes in length could only be accomplished by either the dynamomt t er or the cable acting as a spring (stretching) therefore the change in force would be truch higher. The plant manager disagreed with the inspectors' analysis but agreed that caution was needed. The PRC subsequently approved steps to initially establish a lower value vertical pull (less than 1100 pounds), pull horizontally slowly while monitoring the change in the dynamometer reading, and stop alllifting attempts if the change in force exceeded 10 pour,ds. The actual results showed an increase of about 100 pounds with about a 6-inch horizontal pull. The fuel manufacturer subsequently disassembled and reassembled the two stuck bundles inside the reactor. The operators then transferred the bundles to the SFP without any further difficult The inspectors Jetermined that no violations occurred. As noted in Section E3.1 above, the fuel transfer cask and fuel moves within the reactor or SFP are not classified as

" safety related." At the exit meeting, in response to the inspectors inquiry if the two examples of poor engineering assessments represented overall engineerirsg performance, the plant manager stated that the two assessments had been inadequate, but he did not feel thi.t they represented overall engineering performance. Additionally, the plant manager felt that the engineering staff was heavily involved with overall operation . _ _ _ _ _ _ _ _ __________-____ _ _ -

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l f,onclusions Two engineering analyses supporting defueling were either incomplete or incccurat The fuel bundle removal procedure was properly modified only after the inspectors identified concerns. The inspectors noted a relaxed safety focus as indicated by the licensee accepting two poor engineering assessments, initially directing that a heavy load 1 be left unattended over the reactor, accepting a verbal procedure change, and t. eating dynamometer indications as spurious fluctuations. Overall, despite the flawed analyses i and the relaxed safety focus, the licensee safely defueled the reactor in a cautious and !

methodicel manner with no fuel damag E8 Miscellaneous Engineering issues (92902)

As a minimum, the inspectors reviewed the existing open items under the following criteria: (1) the issue is not applicable to a permanently sr.atdown or decommissioning reactor, (2) the issue does not raise potentially generic concerns, (3) the issue does not involve any pending enforcement action or open investigation, and, (4) the issue does not involve any indication of willful violation E8.1 The licensee's corrective actions for the items listed below were of concern only for power operations and are therefore administratively closed:

[QWLs_eM91low-Up Item 50 155/94018-02: evaluate corrective actions for valve NC-18 failing to operate in automati (Closed) Follow-Up item 50-155/95008-02: battery charger failed to carry load when stack lights were ac'ded.

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(CIposed) Follow-Upfte_n.L50-155/96005_ _0_1: use of portable fans temporary modification, design change, or operator workaroun (Closed) Violation 50-155/97002-03: penetration inspection not performed in accordance with T E Using Inspection Procedure 90712, the inspectors performed an in-office review of licensee event reports (LERs) at the time of issue. Further inspection efforts (onsite) are no longer warranted due to the licensee's permanent fuel removal. The fcilowing LERs are closeJ (Closed) LER 50-155/96004: cleanup system welds not included in the in-service inspection progra (Olosed)MiB_50-155/96007: interpretation of Technical Specification heatup and cooldown limits for reactor scram (Revision 1).

{.Qlosed) LER 50-155/96008: in-service inspection of ASME (American Society of Mechanical Engineers) Code Class 2 not performed per Section X [Qlosed1LER 50-155/910M: demineralized water piping susceptible to over pressurizatio .

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LQQ1pd) LER 50-155/97004: potentialloss of DC power to primary containment spray and liquid poison system IV, Plant Support R1 Radiological Protection and Chemistry Controls (71750)

R1,1 _Qgneral Comments LrlipeptioJJ_qpp_g Using inspection Procedures 71707 and 71750, the inspectors made frequent tours of the radiologically protected area (RPA) and discussed specific radiological controls with the ALARA (as low as reasonably achievable) coordinator and various radiation protection (RP) technicians. The inspectors observed plant conditions and licensee RP practice Observations and Findinas The inspectors noted involvement by RP pcrsonnel in planning and executing work, as reflected in the preparation for the final plant shutdown and decommissioning. As low as reasonably achievable (ALARA) considerations were primary factors in the licensee's decision to decontaminate room 444 (inside containment), allowing storage of the reactor head in a remote location and thus reducing personnel dose during the decommissioning phase. The inspectors also observed that shoring was used to support the reactor head in room 444, instead of the reactor-head stand, which allowed shielding to be placed around the reactor head to further reduce the dose rate. Storing the reactor head in room 444 also allowed moving the reactor-head thermal shield to the main containment floor, a locatior, e at would simplify early disposal of the thermal shield. The thermal shield was bagged in a large, specially-mar._Sctured bag to reduce the possible spread of contamination, a practice developed from lessons learned during previous thermal shield move The inspectors noted that RP technicians were working closely with the operations and maintenance personnel during performance of evolutions in contaminated or high dose areas, especially during defueling operations. The technicians provided close RP oversight includin, instructing personnel to move to low-dose areas if a job was delayed, and monitoring contamination control practices. The inspectors observed RP personnel giving highh> detailed presentations at each pre-job briefing to ensure that all personnel were thoroughly knowledgeable of the radiological conditions, precautions, and lessons leamed from previous evolutions. The ins;.ectors noted that the high level of involvement by RP personnel resulted in the scheduled work being done under the projected personnel dose estimates and without any personnel. contamination events. However, while observed individual RP personnel performance was good, RP practices by others were not as good. During preparations to de-tension the reactor head, the inspectors observed and pointed out to the RP technicians that a worker was lying down in a contaminated area (discussed fully in IR 50-155/97011)

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c. Conclusions On one occasion, radiation protection personnel did not observe a worker lying down in a contaminated area until alerted by the inspectors. However, the technicians overall efforts were responsible for the total person-dose being under the projected dose estimates, in addition, no personnel contaminations occurred during the inspection perio V. Manaaement Meetinas X1 Exit Meetina Sugrn_a3 The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on October 7,1997. The licensee acknowledged the findings presented. The licensee did not identify any of the documents or processes reviewed by the inspectors as proprietar . . _ - _ - _ - _ _ _ _ - - - _ - _ - - _ _ - _ _ - - _

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PARTIAL LIST OF PERSOW CONTACTED 1,icepsee

. K. Powers, General Manager R. Addy, Plant Manager S. Beachum, Engineering Manager G, Boss, Operations Manager l D. Hice, Maintenance Manager J. Rang, Decomm & Business Manager K. Pallagi, Chemistry / Health Physics Manager W. Trubilowicz, Outage / Work Control Manager G. Withrow, Licensing Manager

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INSPECTION PROCEDURES USED IP 37551: Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observations IP 62703: Maintenance Observation IP 64704: Fire Protection Program IP 71707: Plant Operations IP 71750: Plant Support Activities IP 73753: Insennce Inspection IP 83729: Occupational Exposure During Extended Outages IP 83750: Occupational Exposure IP 90712: in Office Review of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92902: Followup - Engineering IP 92903: Followup - Maintenance ITEMS OPENED and CLOSED Closed 155/94018 02 IFl Evaluate Corrective Actions for NC-18 Failing to j

Operate in Automatic 155/EA 95-057-01013 VIO Violation of Technical Specification 11.3.1.4.D for Approximately 78 l Days 155/EA 95-057-01023 VIO Inadequate Training of Personnel Concerning the O-List 155/EA 95-057-01033 VIO Inadequate Instructions for Repair of BS-5761

! 155/EA 95-057-01043 VIO Lack of Procedure for Operating BS-5761 155/EA 95-057-01053 VIO Failure to inspect Work Associated with BS-5761 Repair 155/EA 95-057-01063 VIO Failure to Perform Post-Maintenance Test on BS-5761 155/EA 95-057-01073 VIO Inadequate Corrective Actions to Numerous DP Alarms on BS-5761 During Tests 155/95006 LER Inoperable Emergency-Diesel Generator and Standby-Diesel Generator 155/95007 LER Automatic Reactor Scram During Plant Startup 155/95008 LER Technical Specification Surveillance Requirement Exceeded 155/95008-02 IFl Oattery Charger Failed to Carry Load when Stack Lights were Added 155/95008-03 IFl Multiple Concerns with Spent Fuel Pool Heat Exchanger Cleaning Job 155/95011-02 IFl Effectiveness of Work Control Measures for Fire Protection Expansion Tank 155/95011-03 IFl Use and Control of Caution Cards 155/96002-03 URI Lack of Evaluation of Breakdown in Work Control Process - Wrong Pipe Hanger Cut 155/96002-04 IFl Review of Breaker Testing Procedures and Data Recording Requirements - Molded Case Circuit Breaker Test issues 155/96002-09 URI Breakdown in Work Control on MO-7070

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i 155/96003-02 IFl iHSR Clarification of SFP issues 155/96004 LER Cleanup System Welds not included in the In-Service inspection Program 155/96004-04 IFl Concems Over Use of Contaminated Laundry as a Store Room 155/96005-01 IFl Use of Portable Fans - Temporary Modification, Design Change, or Operator Workaround 155/96005-02 IFI Inspectors Evaluation of Reactor Depressurization System Battery Failure 155/96006-01 URI Weak Turnover During Shifts (Adequacy of Shift Manning and Shift Turnover)

155/96007 LER interpretation of Technical Specification Heatup and Cooldown Limits for Reactor Scram (Revision 1)

155/96008 LER In-service inspection of ASME Code Class 2 not Performed p Section XI 155/96008-01 IFl No. 2 Reactor Protective System Motor Generator Ove Voltage Trip - Review the Root Cause Evaluation 155/96008-02 VIO Inadequate Post-Maintenance Testing on RDS "B" 155/96008-03 IFl Heavy Camponent Storage inside Containment 155/96010-G1 VIO MO-7C62 Closed without SOP-6 Actions Performed 155/96010 03 ViO Failurs to Perform Surveillance 155/96010-05 VIO Fire 3arrier Non-Functional 155/96012 LER Diesel Fuel 92-day Technical Specification Surveillance Requirement Inadvertently Surpassed 155!96012 02 VIO Two Examples of a 10 CFR 50, Appendix B, Criterion V 155/96013 LER Automatic Reactor Scram initiated by Turbine Trip 155/97002-03 VIO Penetration inspection not Performed in Accordance with Technical Specifications 155/97002-05 IFl Material Condition Concerns with Control Rod Drive System 155/97003 LER Demineralized Water Piping Susceptible to Over Pressurization 155/97004 LER Potential Loss of de Power to Primary Containment Spray and Liquid Poison Systems 155/97004-02 URI Protective Devices Installed on DC Breakers Different than the 0;iginal Design 155/97007 01 IFl Corrective Action to be Reviewed on Emergency Lighting Unit Battery Problems 155/97008-01 VIO CRDs Left with High Insertion Speeds after Time Testing 155!97008-02 VIO Containment Equipment Lock inner Door Left Open

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LIST OF ACRONYMS USED

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MAR As Low As Reasonably Achievable '

ASME ' American Society of Mechanical Engineers-ATWS Anticipated Transient Without Scram CFR' Code of Federal Regulations'  :

CR'- . Condition Report CRD Control Rod Drive - '

DP Differential Pressure .

DRP- Division of Reactor Project EDG Emergency Diesel Generator FHSR Final Hazards Summary Report HX Heat Exchanger .

IFl Inspection Followup Item '

IP inspection Procedure IR inspection Report LCO Limiting Condition for Operation LER . Licensee Event Report NRC Nuclear Regulatory Commission Q-List - Quality List PRC Plant Review Committee RDS Reactor Depressurization System RO Reactor Operator RP Radiation Protecti RPA Radiologically Protected Area

SFP Spent Fuel Pool -

SOM Shift Outage Manager SOP Standard Operating Procedure SRO Senior Reactor Operator SS Shift Supervisor TS Technical Specification URI Unresolved item

- VIO Violation WO Work Order .

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