IR 05000482/1987027: Difference between revisions

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{{Adams
{{Adams
| number = ML20236X508
| number = ML20149J716
| issue date = 12/04/1987
| issue date = 02/18/1988
| title = Insp Rept 50-482/87-27 on 871001-30.Six Unresolved Items Identified.Major Areas Inspected:Plant Status,Operational Safety Verification,Monthly Surveillance Observation,Monthly Maint Observation & Physical Security Verification
| title = Forwards Summary of 880111 Enforcement Conference Meeting W/ Util & NRR in Region IV Ofc Re Significant Weaknesses Noted in Insp Repts 50-482/87-27,50-482/87-28 & 50-482/87-31, Including Procedures for Administrative & Mgt Controls
| author name = Bartlett B, Cummins J, Hunnicutt D, Jaudon J, Skow M, Smith W
| author name = Callan L
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =  
| addressee name = Withers B
| addressee affiliation =  
| addressee affiliation = WOLF CREEK NUCLEAR OPERATING CORP.
| docket = 05000482
| docket = 05000482
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-482-87-27, NUDOCS 8712090250
| document report number = EA-87-213, NUDOCS 8802230185
| package number = ML20236X485
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 4
| page count = 21
}}
}}


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l .e- t a FEB 18 !!N Docket: STN 50-482 License No. NPF-42 EA N0. 87-213 Wolf Creek Nuclear Operating Corporation ATTN: Bart D. Withers President and Chief Executive Officer P.O. Box 411 Burlington, Kansas 66839 Gentlemen:
      .
This refers to Enforcement Conference conducted in the NRC Region IV Office on January 11, 1985, with you and other members of your staff, and Region IV and Office of Nuclear Reactor Regulation personnel. This conference was related to the findings of the NRC inspection conducted during the periods October 1 through November 18, 1987, which were documented in NRC Inspection Reports 50-482/87-27, dated December 7,1987,50-482/87-28, dated December 3, 1987, and 50-482/87-31 dated December 21, 1987, respectivel The subjects discussed at this meeting are described in the enclosed Meeting Sumar In accordance with Section 2.790 of the NRC's "Rules of Practice " Part 2, Title 10 Code of Federal Regulations, a copy of this letter will be placed in the NRC's Public Document Roo Should yuu have any questions concerning this matter, we will be pleased to
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discuss them with yo
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7; .    ,  APPENDIX B:
i a   .U.S. NUCLEAR REGULATORY-COMMISSION    !
 
==REGION IV==
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3 NRC' Ir.spectio'n Report: ' .50-462/87-27
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      :0perating Lice 1se:
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NPF-42:
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J Docket:. 50-482
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Licensee: ' Wolf Creek Nuclear: Operating Corporation (WCNOC)  '
I P.O. Cox 411
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Burlington,-Kansas 66839 Facility Name: : Wolf Creek' Generating Station ~(WCGS)1
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Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas Inspection Conducted: October 1-30, 1987-
  : Inspectors / / A/)  1/!3o  :,
g /E.(Cup, mins,;Se6ior Resident; Inspector,.  - Dats-Vperations
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p. 4 rtle.tt;. Resident Reactor.. Inspector, <Date  ,
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    ( Oper tions      ,
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t. S 'ith, Senior Resident Inspector  Date
  # ( Waterfo d 3, Project Section A      i l
Jf  ,  ll 0 E.'Skow, Reactor Inspector,-Test Programs  ' Cato
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Section, Operations-Branch
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Sincerely,
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Approved: (  . /,t-o , hA  Y Date-
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    + L. J. Callan, Director
J./P. Jaudphp Chief, Project Section A    ..
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Division of Reactor Projects Enclosure:
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l Meeting Sumary cc w/ encl:
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Wolf Creek Nuclear Operating Corporation ATTN: Otto Haynard, Manager
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of Licensing P.O. Box 411 Burlington, Kansas 66839 RIV:DRP/A DRP 1 JITapia:gb LJCal'%n ty/88  [/(1/88     -
8802230185 900218
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gDR ADOCK 05000482      \-
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Wolf Creek Nuclear Operating 2 Corporation Wolf Creek Nuclear Operating Corporation ATTN: Gary Boyer, Plant Manager P.O. Box 411 Burlington, Kansas 66839 Kansas Corporation Comission ATTN: Robert O. Elliott, Chief Engineer Fourth Floor, Docking State Office Building Topeka, Kansas 66612-1571 Kansas Radiation Control Program Director bec to DMB 41C01h bec distrib. by RIV:
    -
Myron Karman, ELD, MNBB (1)
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*RRI   R. D. Martin, RA
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*SectionChief(DRP/A) *DRP RPSB-DRSS  J. M. Hinds, RIII
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*RIV File  Callaway, RI!!
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* Project Engineer DRP/A *R. Hall Lisa Shea, RM/ALF  *P. O'Connor, NRR Project Manager
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  *DRS D. Powers
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Dg M' Ht@)icutt, Olief, Test Prograins
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Daye .: .
Sectid6 Operations Branch  -
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i Inspection Summary
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Inspection Conducted October 1-25, 1987 (Report 50 482/87-27)  y Areas Intpected;. Nonroutine, unannounced inspection including plant status,
  . operational safety verification, monthly- surveillance observation <, month)y'
   . maintenance observe. tion, onsite event followup, physical security verification,.
radiological; protection, refueling = activities,~and onsite followup of reportable events which occurred between October-10 and 15, 1987.,
  -Results: Within the'nine areat inspected, six unresolved items are identified
  .in paragraphs 3a, 3b, 3c, 3d, 4,'and 6. Four open items areLidentified in
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ATTACHMENT,
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;    WOLF CREEK NUCLEAR OPERATING CORPORATION   f
 
    ' JANUARY 11, 1988    !
DETAILS Persons Contacted
i MEETING SUMARY Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station License No: NPF-42      f t
  *B. D. Withers, President and CEO E * M. Grant, Vice President, Quality
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  *F. T. Rhodes, Vice President, Nuclear Operations r
Docket No: 50-482
  *G. D. Boyer, Plant Manager
:  SUBJECT: ENFORCEMENT CONFERENCE CONCERNING NRC INSPECTION FINDINGS i   (Inspection Reports 50-482/87-27,50-482/87-28,50-482/87-31) ;
  *0. L. Mayrard, Manager, Licensing
  *C, M. Estes, Superintendent of Operations
   * D. Rich, Superintendent of Maintenance
  * G. Williams, Superintendent of Regulatory, Quality, and Administrative Service *W. J. Rudolph,-QA Manager-WCGS
  *A. A..Freitag, Manager, Nuclear Plant Engineering (NPE), WCGS K. Peterson, Licensing
  *G. Pendergrass, Licensing
  * M. Lindsay, Supervisor,- Quality Systems
   *C. J. Hoch, QA Technologist J. Goode, Licensing Engineer.-
  *V. J. MacTaggart, Supervisor, Results Er.gineering
,  *S. R. Sparks, Licensing Engineer
  *J, C. Hicks, Supervisor, Safety Servicas
  .
  *J. L. Houghton, Operations-Coordinator, Operations
  *R. H. Belote, Manager, Nuclear Safety Engineering C. Fowler, Instrumentation and Control (I&C) Supervisor M. Nichols, Technical Support Superintendent T. Morrill, Health Physics Supervisor J. M. Pippin,' Manager, NPE B. Bergstrom, Acting Manager, NPE Systems-B. McKinney, Superintendent, Test Support J. W . Johnson, Chief of Security W. B. Ward, General Counsel The NRC inspectors also contacted other members of the licensee's staff j during the inspection perio * Denotes those perscrnel in attendance at the exit meeting held on October 30, 1987, Plant Status The plant was shutdown for a refueling cutage during the inspection
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perio . Onsite Event followup During this report period the NRC inspectors performed onsite inspect. ions of the four events discussed below. On Octcber 16, 1987, in response to
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On January 11, 1988, representatives of Wolf Creek Nuclear Operating   i i
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Corporation met in Arlington, Texat with NRC Region IV and NRR personnel to  i discuss the findings documented in the NRC inspection reports dated   i'
 
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- these events, the licensee . suspended refueling outage work and appointed a task force to investigate and evaluate each of the events. The outage
. work was. suspended for approximately one week, and each job was evaluated prior to being restarted to ensure safety. and quality' requirements were being me Two Workers Contaminated
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On October 10, 1937, two workers were contaminated while working on a portable self-contained water processing system that had been provided by e vender (Durate O , The workers were employees of Duratek, and their duties included the operation of the water processing system. The water processing system consisted of five vessels or demineralizers (numbered 0 thru 4) that were mounted on a skid along with the interconnecting hoses and other components, such as valves, that made up the system. The water processing system was set up in the' low level drum storage area on the'2000-foot level of the radwaste building. At the time of the event, water from the plant was not being processed; however, because of previous ,
processirg operations, the media (i.e., resin / charcoal) in some of I the vessels was partially depleted and radioactive. The depleted media was being sluiced out of the No. 2 vessel into the No. 0 vessel so the the No. 2 vessel could be refilled with fresh media. During the sluicing process, the system clogged. In attempting.to unclog the system, the workers disconnected a pressurized hose. As a result, radioactive media (charcoal) was Llewn on the two workers, the low r level storage area, the ceiling of the area, and nver an 18-foot wall into an adjacent corrido Specifically, between 6 a.m., October 10, 1987 cnd approximately 6 p.m. the workers-had performed the following activities:
o Installed two additional vessels in the water processing system and leak tested the system,   I i
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o Loaded activated carbon into the No. 0 vessel, and o Sluice:I depleted mcdia from the No. 4 vessel into the No. O vessel and added new media to the No. 4 vesse Between approximately 6 p.m. ar.d 6:55 p.m., the Duratek workers started sluicing media from.the No. 2 vessel to the No. O vessel. At approximately 6:55 p.m. they exited the ares and notified the radwaste operator that they thcught the water level in ttee floor drain tank receiving the sluice water wcs too high. The radwaste operator put another floor drain tank on-line to receive the sluice water. At 8 p.m. the Duratek workers returned to the area and attempted to restart sluicing the No. 2 vessel; however, the system was clogged. The workers' attempts to unclog the system included disconnecting hoses which opened up the contaminated systen without :
health physics (HP) being informe I
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e  At approximately 10:15'p.m. the swing shift (3 p.m. to 11 p.m.)
 
radwaste operator observed one of the Duratek workers unhooking hoses and allowing water to flow onto the floor and himself. The radwaste  ,
operator also observed that there was some material on the floor that  '
he assumed was new resi At approximately 10:45 p.m. the mid-shift (11 p.m. to 7 a.m.)
 
operator checked in on the Duratek workers and observed water and media on the floor, the Duratek skid, and the Duratek worker The mid-shift radwaste operator stated that it appeared the Duratek workers were trying to clean up the mess. The Duratek operators asked the radwaste operator to get them some towels and decon soap, and said that they would call HP later. The radwaste operator exited the area and notified HP. He then notified the Duratek workers that HP wanted them to stop work and wait on an HP technician. HP personnel subsequently arrived on the scene, stopped all work in the area,'and took the necessary actions to decontaminate the Duratek  i worker The Duratek workers had apparently attempted to clear the clogged line using air and water but were unsuccessful. The' workers stated that they had vented off pressure and verified that the-pressure indication read "0". They said that they had then disconnected the No. 2 vessel fluice line at the sluice manifold. Tne line was pressurized and blew the radioactive media over the workers ond the are Initially it was thought that only the 2.5 lity volume of the disconnected hose blew down; however, the licensee determined by followup investigation that 4.25 cubic foot of media had accumulated'
in the local sump. This indicates that the No. 2 vessel had been pressurized, not isolated, and had apparently blown dow The licensee HP organization normally monitors the opening of a system containing radioactive material and specifies the protective  j clothing and equipment required to be worn by the workers. In this  !
instance, work was performed outside the scope of the existing  l procedures, and the system containing radioactive material was opened  J without the required HP control ]
The licensee had implemented Duratek procedures for operating the  ..
water processing system. However, these procedures did not include instructions for sluicing media between the vessels, nor did the workers' have a procedure for performing maintenance (i.e., opening  I thesystem).
 
Pending further review by the NRC, this matter is considered to be unresoived(482/8727-01). This event was also inspected by au NRC Region IV radiological protection inspector. The results of that  j inspection will be reported in NRC Inspection Report 50-482/87-2 l
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;. Pressurizer Hydrcgen Burn t
l On' October 14, 1987, at 9:15 a.m. the refueling team inside containraent and HP personnel inside containment informed the control room of a loud noise. A welder wat welding on Valve BB V-102, an isolation valve to one of the pressurizer pressure / level instrument sensing-lines. He informad the control room that, when he struck an arc for his second tack he heard a loud rushing sound, which larted 5 or 10 seconds. He also stated that the sheetmetal cover, taped over the hole left when Valve BB PSV-8010B (a pressurizer code safety valvo) was recoved for testina, was blown off by tre bur' ~
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investigation by the lic6nsee doterniined that the only available source of fuel for the burn was hydrogen in the vapor space of the pressurizer. The burn showed up on Control Room Level Recorder BB LR-459 as a change of approximately 13 percent decrease, z. 2 percent )
increase, and then a return to the before burn level values.~ The  '
burn did not show up on the pressurizer pressure recorde d or the rehetor coolant system-(RCS) temperature recorders. HP semphd the area of the code safety and determined that the airborne activity levels had not increased. Plant safety samled the pressurizer for hydrogen after the burn; none was fou :d., T1e licensee's
  , investigation concluded that the probable source.of hydrogen was the void space in the pressurizer dome. When the pressurizer was filled with water to 95 percent in order to push out hydrogen, enough hydrogen apparently remained in the dome area so that when the pressurizer was drained back down the hydrogen level was above the 4 percent required for ignition.. With the upper instrument root valves removed for maintenance and the code safeties being removed for testing apparently enough oxygen was added to support c bur Preliminary analysis by the licensee shows that the pressure pulse inside of the pressurizer was approximately 30 psig. The licensee has inspected pressurizer supports and performed an inspection of the
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spray nozzle through the manway without finding any evidence of damage. The licensee committed that code' safety valve BB PSY-8010A, which had been tested and reinstalled would be removed and reteste EngineeringEvaluationRequest(EER)87-BB-14waswrittentonavdthe verdor evaluate any possible duage to the pressurizer, and the licensee will evaluate the degassing and 'ressurizer p draindown procedures prior to the next refueling outage. Pending the licensee's evaluation concerning possible damage the pressurizer may have sustained and any required corrective action, this will remain anunresolveditem(482/8727-02), Worker Fatally injured    i At 8:37 p.m. on October 14, 1987, an unusual event was declared injured and a fire in the Train "B" 4160 v AC because safeguardsofswitchgear a man being(NB02) roo The man, an electrician, was fatally injured when he came in contact with energized terminals in the NB02 switchgear. The injured man was taken to Coffey County Hospital, where he was pronounced dead. The fire turned out to be
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smoke.-from dameged electrical.equiprent related to the accident. The i  . unusual event ~was-terminated at 9:11 p.m., on.0ctober 14,-1987 Mr, '
fThe .9902 switchgear 'nad been taken out of service and isolated"oyL 3: ,  Clearance Order 87-0676-NB for'a' scheduled outage to clean, inspect,.
1  and test the equipment,
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The 4160 v. AC shfeguards power distribution system had two redundant
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.. buses, NB01 snd NB02. ' Power was supplied to these buses'from two engineered safety.. features (ESF) transformers, XHB01 and XNBC2. ESF k
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Transformer XNB01 supplied normal powar to NB01 thru Circuit ";
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    -Breaker NB0112 and alternate power to NB02 thru. Circuit Breaker NB0212< ESF Transformer XNB02. supplied 40rmal power to NB02 thru y  Circuit Breaker NB0209 and alternate powar to NB01 thru Circuit: ,
Breaker NB0109. These circuit breakers were located-in cubicles in
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the NB, switchgear that they were feeding. . The' NB01 and NB02 swf,tchgear cabinets-are located in separate rooms, and the two s
systems are independent of each. othe For the NB02 maintenance outage, ESF Transformer XNB01 was kept
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energized in order to supply normal power to the.NB01 bus, which wa supplying the ~4160 v AC safety-related loads required by Technical Specification. Having ESF Transformer XNB01 energized also energized the' alternate feed te'the NB02 bus, located in Cubicle NB021 Licensee ' personnel also . decided to keep ESF Transformer XNB02
,  ' energized in' order to' provide an alternate feed to the safety loads on NB01. With ESF. Transformer.XNB02 energized, the normal feed to iD02, located in' Cubic'ie NB0209,.was also energized. The operation t;  . shift supervisor, preparing the clearance order to isolate the.NB02 switchgear, discussed on the telephone with an electricai inainter.cnce supervisor the fact.that both ESF Transformers XND01 and ENB02 would remain energized, during the NB02 outage, and therefore the feedside of two cubicles, NB0212 and NB0209, in the NB02 switchgear would be-
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energized. .However, the electrical maintenance supervisor misunderstood and thought that only the feed side of the NB0212 ,
cubicle would be energized during the NB02 outag Prior to startintj work on the day shift, electrical maintenance
,,  -personnel checked the switchgear with voltage detecting instruments but'fziled to detect that Cubicle NB0209 was energized. " Caution Cubicle Energized" ~iabels were attached to Cubicle NB0212, but nene ware attached to Cubicle NB0209. During the day shift, electricians
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cleaneu and inspected the cubicles and tested the circuit breakers in 9    Cubicle hB02 except for the NB0212 cubicle and breaker. The day L   shift personnel thought that Cubicle NB0209 was completely de-energized and performed the maintenance accordingl i 1 .
The night shift ' f 5 p.m. to 3 'a.m.) relieved the day shif t and .
continued the NB02 maintenance work. The electric 1m who was killed was working'in a potential transformer cabinet which was mounted on i  top of the NB0209 cubicle. The electricians were working on top of L,
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  ' the.NB02. cubicles and had removed the_ covers,from the top of the-
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potential transformer. cabinets. 'With~the top off of NB0209. potential
  ! transformer cabinet, the energized,. stationary disconnect terminals
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that fed the potential transformertwere exposed. It was these-x  ' exposed . terminals that the electrician' contacte . . i , The..NB02 maintenance work was being accomplished ire acccrdence with
  , Procedure MPE E009Q-01, Revision 0, "13.8 KV and 4.16 KV Switchgear
> . Inspection and Testing."
 
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Licensee's Procedure MPE E0090-01 contained a number of steps that should-have identified the presence of high. voltage in the NB02
  . cubicles. -These steps.are discussed below:
o Section_6.0 and 6.1 of MPE E009Q-01 stated:
  " WORK PERFORMANCE INSTRUCTIONS
'
NOTE: Recheck to make sure all the su the switchgear, tie breaker,:(if used)pply
      , high'breakers voltage to supply breaker and space heater breaker, are OPE High voltage breakers should be:in their racked down positio .1' Check the electrical drawings and identify any area (s)
which will.have high_ voltage potential.present even when the Bus'is grounded. List the areas on the L    Attachment "A" Sign-pff Sheet."
 
The only area listed on Attachment "A" was "NB0212-Feeder from p  XNB01."
 
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o Step 6.3.9.3 statedi "Using'high voltage gloves and testor, check the rossettes to ensure that there is no voltage presen Check phase to phase, and phase to ground."
 
The feed side rossettes should have been energized when the check was performed, but the voltage was not detecte o Step 6,3.9.4 stated: "If no voltage was detected in 6.3.9.3, ensure that each of the high voltage connections is discharged."
 
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If an attempt had been made to' discharge the feed side
-  rossettes, the fact that they.were energized would have been detected.
 
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o Step 6.3.9.5 stated: " Clean the insulator and high voltage-connection in each tube. Check the insulators for cracks and
December 7, 3 and 21. The attendance list and licensee presentation are attached. The meeting was held at the request of NRC, Region IV.
  .the ressettes.for damaged fingers."


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Step?6.3 &,5 was signed off for Cubicle'NB0209. This indicated k <
i The NRC discussed its concerns involving the significant weaknesses in the
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  .that people performing this step could hav? been injured, since Steps 6.3.9.3_ and 6.3.S.4 wers apparently not adequately performed.-
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b  'o Section 6.4 of MPE E009Q-01 provided the instructions for performing maintenance on the.pctential transformer o Step 6.4.2 and the note preceding it st?ted:
P  " NOTE:. Use. caution when performing Step-6.4.2. High E  voTtage~ potential may be presen .4.2- Remove bolted panels as neces:;ary to obtain access -
g    to the stationary portion of the high voltage
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disconnects."
 
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        ' Step 6.4.3' stated: "Using the high voltage gloves and testor, check the stationary ' disconnects .for high voltage potential. If no potential is found, check that the high voltage connections are discharged."
 
It was at'this point'in the procedure that the electrician was  l injured...apparently. without the above steps having been adequately:
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performe !,
  'Pending further NRC review of this event and the licensee's task '
forcefindings,'thisisanunresolveditem(482/8727-03).
 
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In addition,'during'the above event, cooling water flow to th j reactor core was lost for'ap3roximately=17 minutes, whenithe reacter
  ' operators s'ecuted' power to tie NB01 bus in response to the rescue-effort. Approximately one-third of the fuel assemblies were still in ,
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the reactor vessel at the time and the refueling cavity was flooded )
up to greater than 23 feet above the reactor vessel flange.for the -l fuel transfer. This event did not result in any danger to the  :
        '
reactor plant.
 
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        -l Engineered Safety Features Actuations  i I
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On October IS, 1987, degraded voltage on vital DC Buses NK02 and NK04 caused numerous ESF actuations. On Wednesday, October 14, 1987, the j 4160 v AC Bus NB02 was de-energized for routine outage maintenance; this de-energized NG02 and NG04, which are the normal feeds to NK02
  ;and NK04. 'With the normal feeds de-energized, Batteries NK12 and NK14 picked up the loads on Buses NK02 and NK04 The licensee was aware of' the' batteries ~ carrying the load and was monitoring bus
  . voltage regularly in the control room, but no specific gravities were
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taken to monitor. battery capacity; "The licensee anticipated that
;  NB02 would be returned to service prior to the batteries becoming
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licensee's procedures for administrative and managenent controls and the  ;
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i  failure to provide an appropriate level of management oversight during the  s
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recent refueling outage at the Wolf Creek Generating Station. Specifically, L  weaknesses in the area of procedural adherence and quality were discussed. The  j
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licensee discussed the cause for each of the issues identified, how the  i problems were discovered, the safety significance of each matter, and the  !
corrective actions taken by the license I


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depleted. ?However,'because of the electrocution on October 14,L1987 l
,,  '(see paragraph 6.c), Dus NB02 was not returned to service as soon as- !
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was originally ~ expected. The batteries became depleted, and at
% .
  .9 p.m..on October 15, 1987, voltage dropped low enough (85-y DC) on L E  NK04, which. feeds 120'AC Bus NN04 v AC-Bus NN04, that the'following !
  .ESF ectuations were received:   l
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Containment purge isolation'. system actuation 1 '
Fuel building ventilation isolation' system actuation  j
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Control rc% rey.tilation. isolation system _  .;
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  .: One channel ~of auto switchover of the auxiliary feedwater (AFW) !
systems suction to the essential service water-(ESW) system for- H auto switchover.to occur, two required out..of three total !
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Start of motor driven AFW pump "A", and  i L
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Loss of coolant accident and shutdown; sequencer actuatio ,
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Thecontrol' room.operatorsplacedthe.switchforthemotordhiven  l ('
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  . auxil.iary feed pump "A" in pull _-to-lock position,' suspended fuel movement and determined the cause of the ESF-actuations. After it j
        ! .was determined that depletion of the batteries. caused the'ESF t  1 c  'actuations, the operators started the documentation required in' order ;
to' supply alternate AC power to' Inverter NK24'to restore power to
,
120 v DC Bus' NN04. At- 11:32 p.m.,: prior to the temporary
'
modification being implemented,t Battery NK12 became depleted. This caused actuation ofithose logic circuits needing 2 out of 4 inputs ,
and gave the following:results: ,
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ESF Pump "A" auto started,'and the AFW systems automatically l switched over to ESW systems as the water sourc '
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The 4160 v AC vital Bus NB01 de-energized onLan indicated undervoltage condition (no undervoltage in fact existed).


All "A" train' equipment automatically stripped off of Bus NB01;
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this caused a loss of residual heat removal snutdown cooling to the core.


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Diesel Generator "A" automatically started and closed onto y    Bus NB01, restoring power to the bus.
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Residual heat removal pump *A" and spent fuel pool cooling pump
    "A" were manually restarted; shutdown cooling was lost for approximately 30 seconds, but because of the large volume of
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water over the reactor core, there was apparently no obvious temperature ris +
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Until temporary power was supplied to either NK02 or NK04, the equipment had to be kept in the actuated condition. Approximately 3 hours after the loss of Bus NK12, while investigating an unexplained increase in condensete storage tank level, the operators observed steem generator wet lay-up. levels greater than 25 inches, which b a indication pegged high and a steam generator pressure of 100 psig which corresponds to the discharge pressure of the emergency service water pum It was'at this time that the operators discovered that the manual isolation valves for emergency service water .to auxilliary feedwater, previously thought to be closed, had been reopened on  .
October 8, 1987. Operator aid magnetic tags on the control board indicated the cross connect valves were shut. Although the auxiliary feedwater pumps were stopped, the pressure in'the essential service
  '
  . water system was.high enough to force water through the AFW pumps and into the minimum flow lines to the condensate storage tank and through the main discharge lines into all four. steam generator Over 3 hours, approximately 10,000 gallons.of lake water had been placed in the condensate storage tank a9d in each of the steam generators. This caused the main steam lines to be completely filled with water. The oparators manually isolated the auxilliary feedwater  .
connect to essential service water. The licensee's investigation of  l
        '
the event has determined:
  *
The steam generator vendor's preliminary analysis of the event has indicated that the steam generators have suffered no damage  j from the poor chemistry of the lake wate i
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Steam generator sludge lancing should be conducted as was  !
originally schedule '
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An analysis of' the batteries for possible damage caused by the deep discharge should be performe Operator aids (magnetic tags) should be controlled to ensure they reflect the true status of equipmen :
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The control work authority for outage related work has been  !
moved to the control room to ensure better communications with the plant operator *
Procedures will be prepared prior to the start of the next refueling outage to allow immedicte connection of all NK buses to alternate power supplies as require Completion of the licensee investigation and implementation of their corrective actions will remain an unresolved item (482/8727-04),
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4. Operational Safety Verification The NRC inspectors verified that the facility is being operated safely and in conformance with regulatory requirements by direct observation of '
licensee facilities, tours of the facility .. interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operations, and reviewing facility record The NRC inspectors, by observation of randomly selected activities and interview of personnel verified that physical security, radiation protection, and fire protection activities were controlle By observing accessible components for correct valve position and electrical breaker position and by observing control room indications, the NRC inspectors confirmed the operability of selected portions of safety-related systems. The NRC inspectors also visually inspected. safety components for leakage, physical damage, and other impairments that could prevent these components from performing their designed safety function Selected NRC inspector observations are discussed below:
o On October 12, 1987, during a routine plant tour, the NRC inspector observed Spool Piece EF-124-HBC-10" removed downstream of Essential Service Water (EF) Valves EF V-048 and EF V-050. No workers were observed in the vicinity and the cleanliness requirements of Plant Administrative Procedures ADM 01-034 and ADM 01-110 were not being complied with in that the open ends of the remaining pipe were not covered.to prevent entry of foreign material. This matter was brought to the licensee's attention and promptly corrected, o On October 13, 1987, the NRC resident inspectors at Callaway informed the Wolf Creek resident inspectors of a problem with the control room ventilation isolation system (CRVIS). As discussed in NRC Inspection Report 50-483/87-23, Callaway identifie' problems with the operability of CRVIS. These identified problems centered on mispositioned, which prevented pressurization of equipment rooms when required. In response to this information, the NRC resident inspectors asked the licensee if the same situation existed at WCNO At the NRC exit meeting held on October 30, 1987, the licensee committed to resolve this item prior to startup from the refueling outage. The licensee's response to this information is not yet completed, and this will be considered an unresolved item (482/8727-05).
5. Monthly Surveillance Observation The NRC inspectors observed selected portions of the performance of surveillance testing and/or reviewed completed surveillance test procedures to verify that surveillance activities were performed in accordance with Technical Specification (TS) requirements and
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administrative procedures. The NRC inspectors considered the following elements while inspecting surveillance activities:
o Testing was being accomplished by qualified personnel in accordance with an approved procedur o The surveillance procedure conformed to TS requirements, o -Required test instrumentation was calibrate o TS limiting conditions for operation (LCO) were satisfie o Test data was accurate and complete. Where appropriate, the NRC inspectors performed independent calculations of selected test data to verify their accurac o The - 4 rmance of the surveillance w edure conformed to applicable acnin arative procedures, o The surveillance was performed within the required frequency and the test results met the required limit Surveillance witnessed and/or reviewed by the NRC inspectors are listed below:
STS MT-039, Revision 2, " Chemical and Volume Control System Relief Valve Testing," performed on October 6, 1987 STS IC-505A, Revision 3, " Calibration of Steam Generator Narrow-Range Level Transmitters," performed on October 7, 1987 STS 10-926, Revision 1, " Surveillance Test Component Cooling Water System Automatic Valve Actuation," performed on October 7,1987 STS MT-027, Revision 3, " Snubber Functional Test," performed on October 7 and 14, 1987 STS MT-042, Revision 2, " Containment Spray System Relief Valve Test,"
;
performed on October 8, 1987 No violations or deviations were identifie . Monthly Maintenance Observation The NRC inspector observed maintenance activities performed on safety-related systems and components to verify that these activities were conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The
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following elements were considered by the NRC inspector during the observation and/or review of the maintenance activities:
o  .LCOs were met and, where applicable, redundant components were operable, o  Activities complied with adequate administrative control .
o  Where required, adequate, approved, and up-to-date procedures were used, o  Craftsmen were qualified to accomplish the designated task and technical expertise (i.e., engineering, health physics, operations)
was made available when appropriat I o  Replacement parts and materials bcing used were properly certifie o  Required radiological cortrols were implemente o  Fire prevention controls were implemented where appropriat ,
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o  Required alignments and surveillance to verify post maintenance operability were performe o  Quality control hold points and/or checklists were used when appropriate and quality control personnel observed designated work .,-
activitie Selected requests p(WR) listed below were observed and related documentationortions of the mainten reviewed by the NRC inspector:
N Activity WR 11361-85  EM V8853B SI "B" discharge relief leak WR 91041-87  EJ FCV-611 valve did not reopen automatically WR 60392-87  Hydrogen Mixing Fans CGN03B refueling inspection .
WR 60051-87  3" Pressure Relief Valve BG V-8120, 5 year bench -
test WR 60048-87  6" Pressurizer Safety Valve BB HV-8010A verify setpoints and operability WR 60026-87  Snubbers, Mechanical 1/4/EG-16R502 functional test WR 60056-87  3/4" Pressure Relief Valve EN V-057, 5 year setpoint and operability test  *
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WR 02828-87 EF V-090 replace reducer WR 60049-87 6" Pressurizer Safety Valve BB HV-8010B verify setpoints ari operability WR 60039-87 Hydraulic Snubbers PD 89070-002 steam generator '' .
WR 60238-87 Main Breaker /NG0201, performed maintenance in > '
      ,
  -accordance with Procedure MPE E017Q-04, Revision 4, observed October 14, 1987
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Switchgear Cubicle and Bus /NB02, performed in
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WR 60224-87 accordance with Procedure MPE E008Q-01, Revision 0, observed October 14, 1987
'WR 60229-87 LCFDR NG02 Circuit Breaker /NB0213, performed in accordance with Procedure MPE E009Q-02, Revision 4, observed October 14, 1987 FHP 02-008 Revision 1, " Reactor Cavity Seal Ring Installation and Removal," performed on October 8, 1987 FHP 02-009 Revision 3, " Reactor Vessel Stud Removal, Cleaning, and Installation" HR 00469-87 Underground coat and wrap on Line AL-001-HBD-12" WR 02887-87 Inspect and rework as required, foam penetration closures, Fire Zones C-22, C-33, and C-34
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WR 02939-87 Battery NA-12, clean cell, terminals, replace cell connections, and check intercell resistance Selected NRC inspector observations are discussed below:  ',-
      '
o During this inspection period all pressurizer code safety relief valves were tested to verify their setpoints with the following results:    ,
Valve  As-Found Setpoint BB PSV-8010A  *2390 psig m
Valg  As-Found Setpoint BB PSV-8010B  *2310 psig BB PSV-80100  *2185 psig
  *Should be 2485 pound per square inch gage (psig) plus or minus 1%
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t-c With:the.' pressurizer power operated relief valves set i'or 2335 psig and~ normal plant operating, pressure around 2235 psig, the results of
  .the tests for Valves BB PSV-8010B and BB PSV-8010C'were placed in ^
   . doubt.< To test these valves, the-licensee' removed them from th ~
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pressurizer and'using a supply of pressurized nitrogen connected to the' valve;.i.nlet slowly: increased the pressure.'until the valve lifte The valves were preheated for 8 hours to ensure proper valve bod temperature ' prior to testing. . The licensee:is attempting to. discover the reasonsLfor the low'as-found setpoints. . Testing of these code.;
safeties is' discussed in NRC Inspection Report 50-482/86-34, paragraph 15:(0penItem 482/8634-04). :Atithe NRC exit meeting held-
  'on October-'30,-1987, the licensee committed to resolving this item
  ' prior to startup after the refueling outage. .Pe'nding the licensee' .
,
  . resolution,of the causes of low as-found setpoints ofLthe pressurizer
    .
code safetiet,.this will be considered an unresolved
  . item'(482/8727-06).
 
0: NRC: Inspection Reports 50-462/87-15, paragraph.5 and 50-482/87-20,
_ paragraph 5 discuss. problems that the licensee has been experiencing-with erosion of pipe.: On October 2, 1987, NRC's Office of Naclear
  . Reactor Regulation (NRR) sent the licensee a letter confirming the 3 '
  ,NRC' staff's understanding of the licensee's actions.-
  . For tracking purposes, the' items listed below have; been taken- from the NRR' letter and will be made open items:
a' . . Prior to returning to operation following the 1987. refueling
  . outage, the licensee will complete the ultrasenic inspection of all safety-related piping locations having a high-probability of-erosion /corrosiondamage(0penItem 482/8727-07). Prior to returning to operation following_the 1987 refuelin'g outage, the licensee will replace the. areas where piping-has experienced through wall or below minimum wall thinning with stainless steel components (0 pen Item 482/8727-08).
 
Prior to returning to operation following the 1987-refueling
      ' i outage,.the licensee will. repair er replace all components that are projected to erode / corrode to less than minimum wall'
thickness during the third' fuel cycle (0 pen Item 482/8727-09), By November 30, 1987, the licensee will provide for NRR review a plan and. schedule for corrective measures that will eliminate
  . erosion / corrosion related pipe thinning caused by flow disturbances in' piping systems ~as a result of butterfly valve throttling and piping configurations (0 pen Item 482/8727-10).
 
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Enclosure ,1 Enforcement Conference Attendance List - Region'lV
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Wolf Creek Nuclear Operating Corporation NRC Attendees: . J. Montgomery, Deputy Regional Administrator J. Callan, Director, Division of Reactor Projects B. Beach Deputy Director, Division-of Reactor Projects D. Powers, Enforcement Officer J. Jaudon, Deputy Director, Division of Reactor Safety P. O'Connor, Project Manager, NRR J. Milhoan, Director Division of Reactor Safety B. Bartlett, Senior Resident Inspector J. L. Scott, Enforcement Staff B. Murray, Chief. Facilities Radiation Protection Section Licensee Attendees: B. D. Withers, President, Wolf Creek Nuclear Operating Corporation F. T. Rhodes, Vice President, Nuclear Operations G. D. Boyer, Plant Manager 0. L. Maynard, Licensing Manager Other Attendees: G. L. Koester, Vice President, Kansas Gas and Electric Company C. J. Ross, Director of Power Engineering, Kansas City Power and Light B. Goshorn, Planning Engineer, Kansas Electric power
  ' Physica'l Security Verification The NRC inspectors verified that the facility physical security' plan (PSP)
{  Cooperative
is being complied with by direct observation of licensee facilities and      .
security personne l
  'The NRC inspectors' by . observation of randomly selected activities verified that-search equipment was operable, that the. protected area barriers and      '
vital. area barriers were well. maintained, that access control procedures were followed and that appropriate compensatory measures were used when equipment was inoperabl No violations or deviations were identifie . Radiological Protection By performing the following activities, the NRC inspectors verified that radiologically related activities were controlled. in accordance with the licensee's procedures and regulatory requirements:
o Reviewed documer.ts such as active radiation work permits and the health physics. shift turnover lo o Observed ersonnel activities in the radiologically controlled area (RCA _such as:
    . Use of the required dosimetry equipment,
    . " Frisking out" of the RCA, and
    . Wearing of appropriate anti-contamination clothing where require o Inspected postings of radiation and contaminated area o Discussed activities with radiation workers and health physics supervisor No violations or deviations were identifie . Refueling Activities The NRC inspectors observed refueling activities related to reactor vessel head removal, installation of the reactor cavity seal ring, flooding of refueling pool, and removal of the upper internals. The NRC inspectors also observed fuel handling operations in the reactor and the spent fuel pool area while the fuel assemblies were being transferred from the reactor vessel to the spent fuel pool. The refueling activities were performed in accordance with Licensee Procedures FHP 02-001, Revision 6,
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" Refueling Procedure," and FHP 02-011, Revision 9, " Fuel Shuffle and Position Verification." The NRC inspectors also. verified that operability of refueling-related equipment was periodically verified by the license '10. Followup of Events One objective of this part of the inspection was to followup on short-term correcti"e actions taken by the licensee subsequent to a series of four incidents which occurred between October 10 and 15, 1987, during the second WCGS refueling outag The secon:f objective was to establis confidence, by observation of work in progress and by rcview of work control procedures, that the licensee had scfficient controls over work,
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clearances, procedure development, and that job interface was such that the outage.can proceed in a safe manne The series of four events, as discussed previously in this inspection report, occurred as follows:
On October 10, 1907, aSout 4 cubic feet of radioactive resin was spilled,in an enclosed room in the radioactive waste building when a pressurized hose was disconnected during a resin transfe *
On October 14, 1987, a combustible mixture of gases ignited inside the reactor coolant system pressurizer while welding on a pressurizer level instrument isolation valv On October 10, 1987, an electrician.was fatally electrocuted while cleaning electrical switchgear containing energhed circuit During the evening of October 15 and the morning of October 16, 1987, Engineered Safety Feature (ESF) actuations occurred as a result of decreased battery bus voltage. The voltage degradation reportedly occurred because safety-related electrical Bus NB-02 was deenergized longer than expected for outage work. The delay appeared to be caused, in part, by recovery actions'from the electrocution inciden Because of an improper isolation of the Auxiliary Feedwater System, the ESF actuation resulted in lake water contamination of all four steam generators when the Auxiliary Feedwater System actuate The licensee temporarily suspended the refueling outage, with exception of: (1) completing the removal of three remaining fuel assemblies from the reactor vessel, thus placing the reactor in a completely defueled conditinn, .(2) steam generator cleanup, and (3) getting power back to electrical Bus NB-02. The licensee stated that these activities were l completed to enhance safety and preclude degradation of the secondary I systems. The NRC staff found those actions acceptabl In a meeting in the NRC Region IV office, as stated in WCNOC Letter WM 87-0277 dated October 20, 1987, the licensee described a
< three-phase approach to the resumption of the outage. This was preceded I
by briefings to project personnel on the work stoppage and the formation LE___
 
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of teams to investigate the four_ incidents. Although the events did not, on the' surface, appear to be related, the. licensee committed to review the events together to determine if a common root cause existe Phase 'One of the recovery consisted of resumption of selected jobs, which were evaluated as having little or no potential for causing additional incidents like those. described abov Examples of Phase One recovery activities included testing of water-operated valves, erosion /ccrrosion measurements, fire protection inspection / rework, turbine generator work, service water system work, and shuffling of fuel in the spent fuel pool (but no movement of fuel into the-reactor). Some of these activities were was observed by the NRC inspectors, and it ;did not appea_r that any safety problems would result, provided the work continued to progress in the orderly, controlled manner' observe Phase Two of the recovery resumed a limited amount of additional work after each work group had conducted meetings with employees and contractors emphasizing the need to do the job right, to follow procedures, and to understand fully the job' requirements and safety
  . precautions. Interviews with varicus department supervisors and on the job discussions with selected workers, QC inspectors, and HP technicians led the NRC inspectors to conclude that the meetings had had a positive impact on job performance. The NRC inspectors cautioned licensee management at the co.nclusion of the inspection that the four incidents themselves may have made the strongest impression rather than the meetings. The licensee was encouraged to ensure, through additional actions, that the workers' current resolve to do their jobs correctly and safely does not fade with tim As the jobs under Phase Two resumed, the NRC inspectors observed work in progress, conducted candid discussions with workers, and reviewed the ;
procedures and documents used by the workers. As committed in Letter WM 87-0277, it appeared that the jobs were adequately prebriefed, ,
  " walked-down" to provide added assurance that each job can proceed safely, j and that the procedures were adequate for the circum;tance The NRC inspectors ideatified two areas where, if left unchecked, could result in problems in the future. On October 27, 1987, the NRC inspectors observed two Instrumentation and Control (1&C) technicians performing Surveillance Test Procedure STS-IC-708, which is a time-response test of the Main Steamline Pressure Instrument Channel The NRC inspectors noted that the technicians had checked off a step in the procedure before it was completed. The step directed the reader to perform a series steps in another part of the procedure. When the NRC inspectors questioned the the checking off of a step in the procedure prior to completion, the technicians indicated that occasionally they will check or sign off a step such as this to acknowledge understanding and at other times they sign off to document completion. Although ADM 02-300, ;
i
  " Surveillance Testing," does not specifically address this practice, the )
l  NRC iospectors expressed concern that if a procedure is interrupted, ;
confusion may occur when the task is resumed. Licensee management stated l
 
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y that it was expected that people would only sign or check off steps in a procedure upon completion of the steps, and that the appropriate training would be implemented. The NRC' inspectors brought this to the attention of the resident inspector for routine followu '
A second observation brought to the attention of licensee management was the existence of numerous puddles of water on the floor in radiologically controlled areas, particularly in the containment building. On one hand  1 the NRC inspectors noted emphasis in general employee training on treating I unidentified puddles as potential radioactive spills, but in practice nobody appeared to be overly concerned about avoiding them or getting them cleaned up. The NRC inspectors expressed concern thats as time goes on, more of these puddles may be contaminated and, if ignored, could result in an uncontrolled spread of contamination. The licensee acknowledged this and agreed with this comment. This will be followed up during the  i resident inspector's routine inspection tour On October 30, 1987, the licensee issued a revised outege schedule which reflected a 2-week delay as a result of the work stoppage and, in addition, another week of time inserted into numerous work segments to a allow for walk-downs and a more deliberate work pace. This was an indication of WCN0C management's resolve to complete the refueling outage at a slower pace with more emphasis on planning and safet As part of the inspection interview process the NRC inspectors met with ,
i  and interviewed five operator personnel, including a shift supervisor (SS),aseniorreactoroperator(SR0),tworeactor operators (R0s), and a nonlicensed equipment operator. The nperators all  '
stated that the outage pace had slowed. They appeared to understand the new controls placed on the magnetic " isolated" signs used in the control room. The operators appeared to understand that they had the authority to /
limit the number of personnel and activities taking place in the control I
room. As part of their corrective action, the licensee moved the SS back
'
to an alcove in the control room to perform his outage-related dutie The operators expressed the opinion that this move improved communications between the SS and the other operators. The SS also stated that, as part ,
of the corrective action, he was attending the morning plan of the day '
meetings. The operatoc's opinion was that his attendance at the meetings could improve some perceived problems with seneduling of individual job This, they hoped, would improve their clearance processing work load. in addition, tne NRC inspectors found that, during control room observation, the operators conducted their activities in a professional manne The NRC inspectors reviewed changes made to procedures as a direct result of the incidents discussed abcve, and without benefit of detailed investigation as to causes:    -
ADM 02-100, Revision 15, " Clearance Order Procedure," was changed on October 20, 1987, to require addition of magnetic status tags placed ,
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N I    21 on the main control board.(MCB) to the applicable clearance order to ensure removal of.the tags when the clearance order is closed. These tags were previously not under any administrative control '
ADM 02-110, Revision 9, " Control of Information Tagging," was revised
'
on October 28, 1987, to support the change to ADM 02-100 abov *
MPE E009Q-01, Revision 1, dated October 15, 1987, "13.8.KV and 4.16'KV Switchgear Inspection and Testing"
* ADM 08-201, Revision 3, Temporary Procedure Change'MA 87-410, dated -
October 27, 1987, " Control of Maintenance and Modifications"
*
ADM 01-057,. Revision 12, Temporary Procedure Change MA 87-349, " Work -
Request" There were several changes and added procedures associated with the resin spill. These are discussed in NRC Inspection Report 50-482/87-2 The NRC inspectors clso reviewed the following administrative control procedures, and found no significant problems:
*
ADM 02-021, Revision 9, "Use of Procedures in Operations." This procedure establishes the guidelines for the use of Operation Procedure *
ADM 02-300e Revision 10. " Surveillance Testing." This procedure discusses the conduct of surveillance and the use of surveillance procedure *
ADM 02-100, Revision 15, " Clearance Order Procedure." The purpose of this procedure .is to provide methods to ensure the safety of-personnel and protection of plant equipment during maintenance and operation. It also ensures that safety-related equipment removal from, and restoration to, w rvice is independently verifie ADM 02-110, Revision 9, " Control of Information Tagging." This procedure describes the authorization, documentation, and review required to ensure operator aids and information tags are current, complete, and necessar *
ADM 07-100, Revision 32, " Preparation, Review, Approval, and Distribution of WCGS Procedures." -This procedure delineates the process for preparation, review, approval, and distribution of procedures required by Section G of the Technical Specification At the conclusion of this inspection, the licensee had not entered )
Phase Three of the outage resumption. The licensee committed not to go into full resumption until completian of the investigations into the four
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  ; incidents above,: completion of all inanediate corrective actions' identified as.a result of.the: investigations and management procedure, and-notification to the NRC of-WCNOC's. intention-to-resum .
  .o 3 Fuel. handling operations were conducted in accordance with approved procedure '
o" TechnicalfSpecification. requirements were' me .o- Good housekeeping and loose object control were maintained in the-refueling cavity-and the spent fuel pool area . Licensed'and/or. qualified personnel performed. tasks and manned-stations whers, required by technical specifications or procedure During discussion,-the NRC inspector determined that the licensee had
,
performed a' safety. evaluation in accordance with 10 CFR Part 50.59 for the
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  'new core' loa 'No violations or deviations were identifie i
[ '1 Unresolved Item I  . Unresolved items.are matters aboutlwhich more information;is required in
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orderEto ascertain whether they. are acceptable' item, items of'
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noncompliance, or deviations. Six unresolved items disclosed during the-
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l  . inspection.'are discussed in~ paragraphs 3a, 3b, 3c, 3d, 4,'and P Exit Meeting i'
l:  The NRC inspectors met with' licensee personnel to discuss the scope and L  findings of this inspection.on October:30, 1987.
 
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Revision as of 13:41, 11 December 2021

Forwards Summary of 880111 Enforcement Conference Meeting W/ Util & NRR in Region IV Ofc Re Significant Weaknesses Noted in Insp Repts 50-482/87-27,50-482/87-28 & 50-482/87-31, Including Procedures for Administrative & Mgt Controls
ML20149J716
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 02/18/1988
From: Callan L
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To: Withers B
WOLF CREEK NUCLEAR OPERATING CORP.
References
EA-87-213, NUDOCS 8802230185
Download: ML20149J716 (4)
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l .e- t a FEB 18 !!N Docket: STN 50-482 License No. NPF-42 EA N0.87-213 Wolf Creek Nuclear Operating Corporation ATTN: Bart D. Withers President and Chief Executive Officer P.O. Box 411 Burlington, Kansas 66839 Gentlemen:

This refers to Enforcement Conference conducted in the NRC Region IV Office on January 11, 1985, with you and other members of your staff, and Region IV and Office of Nuclear Reactor Regulation personnel. This conference was related to the findings of the NRC inspection conducted during the periods October 1 through November 18, 1987, which were documented in NRC Inspection Reports 50-482/87-27, dated December 7,1987,50-482/87-28, dated December 3, 1987, and 50-482/87-31 dated December 21, 1987, respectivel The subjects discussed at this meeting are described in the enclosed Meeting Sumar In accordance with Section 2.790 of the NRC's "Rules of Practice " Part 2, Title 10 Code of Federal Regulations, a copy of this letter will be placed in the NRC's Public Document Roo Should yuu have any questions concerning this matter, we will be pleased to

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discuss them with yo

Sincerely,

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+ L. J. Callan, Director

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Division of Reactor Projects Enclosure:

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Wolf Creek Nuclear Operating Corporation ATTN: Otto Haynard, Manager

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of Licensing P.O. Box 411 Burlington, Kansas 66839 RIV:DRP/A DRP 1 JITapia:gb LJCal'%n ty/88 [/(1/88 -

8802230185 900218

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gDR ADOCK 05000482 \-

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Wolf Creek Nuclear Operating 2 Corporation Wolf Creek Nuclear Operating Corporation ATTN: Gary Boyer, Plant Manager P.O. Box 411 Burlington, Kansas 66839 Kansas Corporation Comission ATTN: Robert O. Elliott, Chief Engineer Fourth Floor, Docking State Office Building Topeka, Kansas 66612-1571 Kansas Radiation Control Program Director bec to DMB 41C01h bec distrib. by RIV:

Myron Karman, ELD, MNBB (1)

  • RRI R. D. Martin, RA
  • SectionChief(DRP/A) *DRP RPSB-DRSS J. M. Hinds, RIII
  • RIV File Callaway, RI!!
  • Project Engineer DRP/A *R. Hall Lisa Shea, RM/ALF *P. O'Connor, NRR Project Manager
  • DRS D. Powers
  • w/766 I

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ATTACHMENT,

WOLF CREEK NUCLEAR OPERATING CORPORATION f

' JANUARY 11, 1988  !

i MEETING SUMARY Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station License No: NPF-42 f t

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Docket No: 50-482

SUBJECT: ENFORCEMENT CONFERENCE CONCERNING NRC INSPECTION FINDINGS i (Inspection Reports 50-482/87-27,50-482/87-28,50-482/87-31)  ;

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On January 11, 1988, representatives of Wolf Creek Nuclear Operating i i

Corporation met in Arlington, Texat with NRC Region IV and NRR personnel to i discuss the findings documented in the NRC inspection reports dated i'

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December 7, 3 and 21. The attendance list and licensee presentation are attached. The meeting was held at the request of NRC, Region IV.

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i The NRC discussed its concerns involving the significant weaknesses in the

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licensee's procedures for administrative and managenent controls and the  ;

i failure to provide an appropriate level of management oversight during the s

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recent refueling outage at the Wolf Creek Generating Station. Specifically, L weaknesses in the area of procedural adherence and quality were discussed. The j

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licensee discussed the cause for each of the issues identified, how the i problems were discovered, the safety significance of each matter, and the  !

corrective actions taken by the license I

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Enclosure ,1 Enforcement Conference Attendance List - Region'lV

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Wolf Creek Nuclear Operating Corporation NRC Attendees: . J. Montgomery, Deputy Regional Administrator J. Callan, Director, Division of Reactor Projects B. Beach Deputy Director, Division-of Reactor Projects D. Powers, Enforcement Officer J. Jaudon, Deputy Director, Division of Reactor Safety P. O'Connor, Project Manager, NRR J. Milhoan, Director Division of Reactor Safety B. Bartlett, Senior Resident Inspector J. L. Scott, Enforcement Staff B. Murray, Chief. Facilities Radiation Protection Section Licensee Attendees: B. D. Withers, President, Wolf Creek Nuclear Operating Corporation F. T. Rhodes, Vice President, Nuclear Operations G. D. Boyer, Plant Manager 0. L. Maynard, Licensing Manager Other Attendees: G. L. Koester, Vice President, Kansas Gas and Electric Company C. J. Ross, Director of Power Engineering, Kansas City Power and Light B. Goshorn, Planning Engineer, Kansas Electric power

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